Uncertainties in Predicted

Isotopic Compositions for

High Burnup PWR Spent

Nuclear Fuel

Office of Nuclear Regulatory Research 

NUREG/CR-7012

ORNL/TM-2010/41
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Uncertainties in Predicted

Isotopic Compositions for

High Burnup PWR Spent

Nuclear Fuel 

Manuscript Completed: September 2010 

Date Published: January 2011 

Prepared by

I.C. Gauld, G. Ilas, and G. Radulescu 

Oak Ridge National Laboratory

Managed by UT-Battelle, LLC

Oak Ridge, TN 37831-6170

M. Aissa, NRC Project Manager

NRC Job Code N6540 

Office of Nuclear Regulatory Research 

NUREG/CR-7012

ORNL/TM-2010/41

iii 

ABSTRACT 

Experimental isotopic assay data for 51 spent fuel samples acquired from domestic and international 

programs have been compiled to provide a database to validate computational predictions of spent fuel 

isotopic compositions important to criticality safety (burnup credit), reactor physics, spent fuel storage 

and transportation, and waste management applications. The data were selected on the basis that they 

include extensive actinide and fission product isotopic assay measurements. The experimental data were 

acquired for fuels irradiated in six different pressurized water reactors: the Three Mile Island 1 and 

Calvert Cliffs 1 reactors operated in the United States, the Takahama-3 reactor in Japan, the Gösgen 

reactor in Switzerland, the GKN II reactor in Germany, and the Vandellós II reactor in Spain. The fuel 

samples cover enrichments from 2.6 to 4.7 wt % 235U, and a wide burnup range, from 9 to 78 GWd/MTU. 

In this report, spent fuel isotopic compositions calculated using two-dimensional isotope depletion models 

and ENDF/B-V–based cross section libraries in the SCALE 5 code system are benchmarked against the 

experimental isotopic data to validate the code and nuclear data libraries and provide estimates of isotopic 

bias and uncertainty for nuclides of highest importance to safety and licensing applications. The 

procedures used to evaluate the experimental data and assess the results are described. 
iv 
v 

TABLE OF CONTENTS 

Page

ABSTRACT ................................................................................................................................................. iii

LIST OF FIGURES .................................................................................................................................... vii

LIST OF TABLES ....................................................................................................................................... xi

ACKNOWLEDGMENTS ......................................................................................................................... xiii

ACRONYMS AND ABBREVIATIONS ............ .................................................................................. xv

1 INTRODUCTION .................................................................................................................................. 1

2 OVERVIEW OF EXPERIMENTAL PROGRAMS AND FUEL CHARACTERISTICS .................... 5

2.1 CALVERT CLIFFS 1 ................................................................................................................... 5

2.2 TAKAHAMA 3 ............................................................................................................................. 7

2.3 THREE MILE ISLAND 1 ............................................................................................................. 9

2.4 GÖSGEN ..................................................................................................................................... 11

2.5 GKN II ........................................................................................................................................ 13

2.6 VANDELLÓS II ......................................................................................................................... 13

3 MEASUREMENT METHODS AND UNCERTAINTIES ................................................................. 17

3.1 PNNL MEASUREMENTS ......................................................................................................... 17

3.2 KRI MEASUREMENTS ............................................................................................................ 20

3.3 JAERI MEASUREMENTS ........................................................................................................ 20

3.4 GE-VNC MEASUREMENTS .................................................................................................... 21

3.5 ANL MEASUREMENTS ........................................................................................................... 21

3.6 STUDSVIK NUCLEAR AB MEASUREMENTS ..................................................................... 22

3.7 SCK•CEN MEASUREMENTS .................................................................................................. 23

3.8 ITU MEASUREMENTS ............................................................................................................. 23

3.9 PSI MEASUREMENTS ............................................................................................................. 23

3.10 CEA MEASUREMENTS ........................................................................................................... 24

4 COMPUTATIONAL ANALYSIS METHODS ................................................................................... 25

4.1 TRITON/NEWT .......................................................................................................................... 25

4.1.1 Computational Models ...................................................................................................... 25

4.2 CROSS-SECTION LIBRARIES ................................................................................................ 29

4.3 RESONANCE PROCESSING ................................................................................................... 29

4.4 ISOTOPIC DEPLETION CALCULATIONS ............................................................................ 30

5 ANALYSIS RESULTS AND DISCUSSION ...................................................................................... 31
vi 

5.1 IMPORTANT ISOTOPES IN SPENT NUCLEAR FUEL SAFETY ANALYSES ................... 31

5.2 COMPARISON OF CALCULATED AND EXPERIMENTAL RESULTS .............................. 35

5.3 DATA EVALUATION CRITERIA ........................................................................................... 44

5.4 STATISTICAL ANALYSES ...................................................................................................... 45

5.4.1 Normality Tests ................................................................................................................. 45

5.4.2 Linear Regression Analysis .............................................................................................. 46

5.4.3 Bias and Uncertainty Analysis .......................................................................................... 46

5.4.4 Confidence Intervals ......................................................................................................... 47

5.5 REVIEW OF LABORATORY MEASUREMENTS ................................................................. 52

5.5.1 PNNL-KRI ........................................................................................................................ 52

5.5.2 JAERI ............................................................................................................................... 52

5.5.3 ANL .................................................................................................................................. 53

5.5.4 GE-VNC ........................................................................................................................... 56

5.5.5 SCK•CEN ......................................................................................................................... 56

5.5.6 ITU .................................................................................................................................... 56

5.5.7 Studsvik Nuclear AB ........................................................................................................ 57

5.6 ANALYSIS OF SELECTED ISOTOPIC DATA ....................................................................... 57

5.6.1 Uranium Isotopes .............................................................................................................. 57

5.6.2 Plutonium Isotopes ........................................................................................................... 59

5.6.3 Neptunium-237 ................................................................................................................. 59

5.6.4 Americium ........................................................................................................................ 60

5.6.5 Curium Isotopes ................................................................................................................ 61

5.6.6 Lanthanides ....................................................................................................................... 61

5.6.7 Cesium Isotopes ................................................................................................................ 63

5.6.8 Metallic Fission Product Isotopes ..................................................................................... 64

5.6.9 Isotopic Bias and Uncertainty Results .............................................................................. 66

5.7 OBSERVATIONS AND RECOMMENDATIONS ................................................................... 68

6 CONCLUSIONS .................................................................................................................................. 71

7 REFERENCES ..................................................................................................................................... 73

APPENDIX A ......................................................................................................... A-1
vii 

LIST OF FIGURES 

Page

Figure 2.1. Layout for Calvert Cliffs 1 assembly D047. .......................................................................... 7

Figure 2.2. Layout for Takahama-3 assemblies NT3G23 and NT3G24 and sample locations. ............... 8

Figure 2.3. Layout for TMI-1 assembly NJ05YU (rod H6) and NJ070G (rods O1, O12, O13). ........... 10

Figure 2.4. Assembly layout and positions of Gösgen ARIANE and MALIBU samples. ..................... 12

Figure 2.5. Assembly layout for GKN II (REBUS) sample rod M11. ................................................... 14

Figure 2.6. Assembly layout for Vandellós II samples. ......................................................................... 15

Figure 4.1. Illustration of the NEWT model used for Calvert Cliffs assembly D047. ........................... 26

Figure 4.2. TRITON assembly model for Vandellós II samples from rod WZR0058 (cycles 7–9). ..... 27

Figure 4.3. TRITON assembly model for Vandellós II samples from rod WZR0058 (cycle 10). ......... 28

Figure 4.4. TRITON assembly model for Vandellós II samples from rod WZR0058 (cycle 11). ......... 29

Figure 5.1. Relative isotopic importance in burnup credit criticality calculations. ................................ 33

Figure 5.2. Relative isotopic contribution to the total decay heat for typical high burnup fuel. ............ 34

Figure 5.3. 235U concentration for TMI-1 samples measured at ANL (Phase I and II) shown as a 

function of (a) estimated burnup and (b) axial height of the sample in the fuel rod ............ 55

Figure 5.4. Comparison of measurements and calculations for 243Am ................................................... 61

Figure 5.5. Comparison of measurements and calculations for 151Eu .................................................... 64

Figure A.1. Comparison of measurements and calculated values for 234U. ........................................... A-2

Figure A.2. Comparison of measurements and calculated values for 235U. ........................................... A-3

Figure A.3. Comparison of measurements and calculated values for 236U. ........................................... A-4

Figure A.4. Comparison of measurements and calculated values for 238U. ........................................... A-5

Figure A.5. Comparison of measurements and calculated values for 238Pu. ......................................... A-6

Figure A.6. Comparison of measurements and calculated values for 239Pu. ......................................... A-7

Figure A.7. Comparison of measurements and calculated values for 240Pu. ......................................... A-8

Figure A.8. Comparison of measurements and calculated values for 241Pu. ......................................... A-9

Figure A.9. Comparison of measurements and calculated values for 242Pu. ....................................... A-10

Figure A.10. Comparison of measurements and calculated values for 237Np. ....................................... A-11

Figure A.11. Comparison of measurements and calculated values for 241Am....................................... A-12

Figure A.12. Comparison of measurements and calculated values for 242mAm. .................................... A-13

Figure A.13. Comparison of measurements and calculated values for 243Am....................................... A-14

Figure A.14. Comparison of measurements and calculated values for 242Cm. ...................................... A-15

Figure A.15. Comparison of measurements and calculated values for 243Cm. ...................................... A-16
viii 

LIST OF FIGURES (continued) 

Page

Figure A.16. Comparison of measurements and calculated values for 244Cm. ...................................... A-17

Figure A.17. Comparison of measurements and calculated values for 245Cm. ...................................... A-18

Figure A.18. Comparison of measurements and calculated values for 246Cm. ...................................... A-19

Figure A.19. Comparison of measurements and calculated values for 247Cm. ...................................... A-20

Figure A.20. Comparison of measurements and calculated values for 133Cs. ....................................... A-21

Figure A.21. Comparison of measurements and calculated values for 134Cs. ....................................... A-22

Figure A.22. Comparison of measurements and calculated values for 135Cs. ....................................... A-23

Figure A.23. Comparison of measurements and calculated values for 137Cs. ....................................... A-24

Figure A.24. Comparison of measurements and calculated values for 140Ce. ....................................... A-25

Figure A.25. Comparison of measurements and calculated values for 142Ce. ....................................... A-26

Figure A.26. Comparison of measurements and calculated values for 144Ce. ....................................... A-27

Figure A.27. Comparison of measurements and calculated values for 142Nd. ....................................... A-28

Figure A.28. Comparison of measurements and calculated values for 143Nd. ....................................... A-29

Figure A.29. Comparison of measurements and calculated values for 144Nd. ....................................... A-30

Figure A.30. Comparison of measurements and calculated values for 145Nd. ....................................... A-31

Figure A.31. Comparison of measurements and calculated values for 146Nd. ....................................... A-32

Figure A.32. Comparison of measurements and calculated values for 148Nd. ....................................... A-33

Figure A.33. Comparison of measurements and calculated values for 150Nd. ....................................... A-34

Figure A.34. Comparison of measurements and calculated values for 147Pm. ...................................... A-35

Figure A.35. Comparison of measurements and calculated values for 147Sm. ...................................... A-36

Figure A.36. Comparison of measurements and calculated values for 148Sm. ...................................... A-37

Figure A.37. Comparison of measurements and calculated values for 149Sm. ...................................... A-38

Figure A.38. Comparison of measurements and calculated values for 150Sm. ...................................... A-39

Figure A.39. Comparison of measurements and calculated values for 151Sm. ...................................... A-40

Figure A.40. Comparison of measurements and calculated values for 152Sm. ...................................... A-41

Figure A.41. Comparison of measurements and calculated values for 154Sm. ...................................... A-42

Figure A.42. Comparison of measurements and calculated values for 151Eu. ....................................... A-43

Figure A.43. Comparison of measurements and calculated values for 153Eu. ....................................... A-44

Figure A.44. Comparison of measurements and calculated values for 154Eu. ....................................... A-45

Figure A.45. Comparison of measurements and calculated values for 155Eu. ....................................... A-46

Figure A.46. Comparison of measurements and calculated values for 154Gd. ....................................... A-47
ix 

LIST OF FIGURES (continued) 

Page

Figure A.47. Comparison of measurements and calculated values for 155Gd. ....................................... A-48

Figure A.48. Comparison of measurements and calculated values for 156Gd. ....................................... A-49

Figure A.49. Comparison of measurements and calculated values for 158Gd. ....................................... A-50

Figure A.50. Comparison of measurements and calculated values for 160Gd. ....................................... A-51

Figure A.51. Comparison of measurements and calculated values for 90Sr. ......................................... A-52

Figure A.52. Comparison of measurements and calculated values for 95Mo. ....................................... A-53

Figure A.53. Comparison of measurements and calculated values for 99Tc. ......................................... A-54

Figure A.54. Comparison of measurements and calculated values for 101Ru. ....................................... A-55

Figure A.55. Comparison of measurements and calculated values for 106Ru. ....................................... A-56

Figure A.56. Comparison of measurements and calculated values for 103Rh. ....................................... A-57

Figure A.57. Comparison of measurements and calculated values for 125Sb. ....................................... A-58

Figure A.58. Comparison of measurements and calculated values for 109Ag. ....................................... A-59
x 
xi 

LIST OF TABLES 

Page

Table 1.1. Summary of evaluated spent fuel measurements ................................................................... 3

Table 2.1. Summary of evaluated spent fuel samples ............................................................................. 6

Table 3.1. Summary of actinide measurement methods and estimated uncertainties ........................... 18

Table 3.2. Summary of fission product measurement methods and uncertainties ................................ 19

Table 5.1. Important nuclides in spent fuel safety applications ............................................................ 32

Table 5.2. C/E-1 (%) for Calvert Cliffs samples ................................................................................... 36

Table 5.3. C/E-1 (%) for Takahama-3 samples ..................................................................................... 37

Table 5.4. C/E-1 (%) for TMI-1 samples measured at GE-VNC .......................................................... 39

Table 5.5. C/E-1 (%) for TMI-1 samples from assembly NJ05YU measured at ANL ......................... 40

Table 5.6. C/E-1 (%) for Gösgen and GKN II samples ........................................................................ 41

Table 5.7. C/E-1 (%) for Vandellós II samples ..................................................................................... 43

Table 5.8. Normality test results for actinides (U, Pu, Am, Np, Cm) ................................................... 48

Table 5.9. Normality test results for fission products (Cs, Ce, Nd) ...................................................... 48

Table 5.10. Normality test results for fission products (Sm, Eu, Gd) ..................................................... 48

Table 5.11. Normality test results for other fission products (metallics) ................................................ 49

Table 5.12. Calculated-to-experimental differences (%)—Major actinides ........................................... 49

Table 5.13. Calculated-to-experimental differences (%)—Minor actinides ........................................... 49

Table 5.14. Calculated-to-experimental differences (%)—Fission products (Cs, Ce, Nd) ..................... 50

Table 5.15. Calculated-to-experimental differences (%)—Fission products (Pm, Sm) .......................... 50

Table 5.16. Calculated-to-experimental differences (%)—Fission products (Eu, Gd) ........................... 51

Table 5.17. Calculated-to-experimental differences (%)—Fission products (metallics) ........................ 51

Table 5.18. Summary of analysis results for all samples used in this study ........................................... 67
xii 
xiii 

ACKNOWLEDGMENTS 

This work was performed under contract with the U.S. Nuclear Regulatory Commission Office of Nuclear 

Regulatory Research. The authors acknowledge the helpful review comments provided by staff of the 

Division of Spent Fuel Storage and Transportation, and guidance from the NRC Project Manager 

M. Aissa. Review of the manuscript by S. M. Bowman, F. Peretz, J. Dole, A. Boone, P. Henson and 

preparation of the final report by D. Weaver are very much appreciated. 
xiv 
xv 

ACRONYMS AND ABBREVIATIONS 

ANL Argonne National Laboratory 

ARIANE Actinides Research In A Nuclear Element 

ATM approved testing material 

BPR burnable poison rods 

CE Combustion Engineering 

CEA Commissariat à l'Énergie Atomique 

C/E calculated-to-experimental 

CSN Consejo de Seguridad Nuclear 

DOE U.S. Department of Energy 

ENRESA Empresa Nacional de Residuo Radioactivo 

ENUSA Empresa Nacional del Uranio, S. A. 

EPRI Electric Power Research Institute 

GE-VNC General Electric Vallecitos Nuclear Center 

GKN II Gemeinschaftskernkraftwerk Neckar Unit II 

ICPMS inductively coupled plasma mass spectrometry 

ID isotope dilution 

IDMS isotope dilution mass spectrometry 

ITU Institute for Transuranium Elements 

JAERI Japan Atomic Energy Research Institute 

KRI Khlopin Radium Institute 

LA luminescent analysis 

LWR light water reactor 

MALIBU MOX and UOX LWR Fuels Irradiated to High Burnup 

MOX mixed oxide 

MS mass spectrometry 

NEA Nuclear Energy Agency 

NRC U.S. Nuclear Regulatory Commission 

OCRWM Office of Civilian Radioactive Waste Management 

OECD Organisation for Economic Co-operation and Development 

ORNL Oak Ridge National Laboratory 

PNNL Pacific Northwest National Laboratory 

PSI Paul Scherrer Institute 

PWR pressurized water reactor 

REBUS Reactivity Tests for a Direct Evaluation of the Burnup Credit on Selected 

irradiated LWR fuel bundles 

RSD relative standard deviation 

SCALE Standardized Computer Analyses for Licensing Evaluation 

SCK•CEN Studiecentrum voor Kernenergie - Centre d'Étude de l'Énergie Nucléaire 

SFCOMPO Spent Fuel Isotopic Composition Database 

TIMS thermal ionization mass spectrometry 

TMI-1 Three Mile Island Unit 1 

UOX uranium oxide 

YMP Yucca Mountain Project 
xvi 
1 

1 INTRODUCTION 

Analysis and evaluation of the bias and uncertainties in the predicted isotopic composition of spent 

nuclear fuel are essential for evaluating the performance and validating the accuracy of computer codes 

and nuclear data used for light water reactor (LWR) safety and licensing calculations. Bias and 

uncertainty in predicted spent fuel isotopic compositions are important in developing safety margins for 

uncertainty, required in criticality calculations that utilize burnup credit (reduction of reactivity associated 

with fuel irradiation). Isotopic uncertainties are also an important source of uncertainty in calculated 

radionuclide activities that define the radiation sources used in postulated severe accident analysis, and 

dose rates and decay heat generation rates for spent fuel handling, transportation, and storage facilities. 

The trend in the nuclear industry towards the use of higher fuel enrichments, advanced fuel designs, 

optimized fuel management strategies, and increasing capacity factors has resulted in progressively higher 

burnup fuel being discharged from commercial LWR plants in the United States. Assemblies now attain 

average burnups of 60 GWd/MTU, with local fuel rod burnup values achieving even higher values. The 

majority of measured isotopic assay data that are currently publicly available to validate code accuracy 

involve fuel with burnups less than 40 GWd/MTU and enrichments below 4 wt % 235U, severely limiting 

the ability to validate computer code predictions for high burnup fuel. Thus, fuel currently being 

discharged from nuclear reactors has characteristics that extend well beyond regimes where experimental 

validation data are readily available. 

Under a project sponsored by the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear 

Regulatory Research, experimental isotopic assay data representative of modern fuel designs and burnup 

characteristics have been acquired and documented in a series of reports for the purpose of code 

validation, and the spent fuel measurements have been applied to validate isotopic calculations performed 

using the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system:1

 

• NUREG/CR-6968 – Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic 

Validation—Calvert Cliffs, Takahama, and Three Mile Island Reactors; 2

• NUREG/CR-6969 – Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic 

Validation—ARIANE and REBUS Programs (UO2 Fuel); 3

• NUREG/CR-7013 – Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic 

Validation—Vandellós II Reactor; 4

• Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation–MALIBU 

Program (UO2 Fuel). The MALIBU program data are currently commercially protected, and details 

of the experiments are not yet available to the public.5

The experimental data analyzed in the present study were limited to include only pressurized water 

reactor (PWR) fuels. The fuel samples were selected on the basis that they cover a relatively wide burnup 

range and include both extensive actinide and fission product measurements. Isotopic data for 51 UO2

fuel samples from commercial assemblies were acquired from domestic and international programs, and 

measurements for more than 60 isotopes are evaluated. The selected samples were obtained from fuel 

rods with initial fuel enrichments from 2.6 to 4.7 wt % 235U and cover a large burnup range from 8 to 

79 GWd/MTU. A summary of the experimental programs and spent fuel samples evaluated in this report 

is listed in Table 1.1. 
2 

This report was prepared to compile and evaluate isotopic assay measurements from different 

experiments into a single document as part of a comprehensive computational validation exercise. The 

present data established an experimental database that can be used to develop guidance on isotopic bias 

and uncertainties, particularly as they apply to applications involving high burnup spent fuel. The 

isotopic uncertainties are selectively evaluated for nuclides of high importance to burnup credit, nuclear 

criticality safety, radiation source terms (neutron and gamma ray sources), and decay heat applications. 

The report is organized into the following sections: 

• Section 2 gives an overview of the analyzed fuel samples considered and the experimental programs 

from which the measurement data were acquired. 

• Section 3 reviews the measurement techniques and uncertainties. 

• Section 4 describes the computational analysis methods and models. 

• Section 5 compares the measured and calculated isotopic results, describes the criteria used to 

evaluate the different data sets, and discusses the results of the evaluation. 

• Section 6 presents conclusions. 

Sections 2–4 present an overview of the experimental programs and the analysis results. A more detailed 

description of the experiments and the fuel design and reactor operating information required to perform 

benchmark code calculations can be found in the separate reports for each program. 
Table 1.1. Summary of evaluated spent fuel measurements 

Reactor 

(country) 

Measurement 

laboratorya

Experimental 

programa

Assembly

design 

Enrichment 

(wt % 235U) 

No. of 

samples

Measuremen

methodsb

TMI-1 (USA) 

ANL 

(USA) 

OCRWM 

(YMP) 

15 × 15 4.013 11 ICPMS, 

α-spec, γ-spe

TMI-1 (USA) 

GE-VNC 

(USA) 

OCRWM 

(YMP) 

15 × 15 4.657 8 TIMS, 

α-spec, γ-spe

Calvert Cliffs 1 

(USA) 

PNNL, KRI 

(USA, Russia) 

ATM 14 × 14 3.038 3 IDMS, LA, 

α-spec, γ-spe

Takahama 3 

(Japan) 

JAERI 

(Japan) 

JAERI 17 × 17 2.63, 4.11 16 IDMS, 

α-spec, γ-spe

Gösgen 

(Switzerland) 

SCK•CEN, ITU 

(Belgium, Germany) 

ARIANE 15 × 15 3.5, 4.1 3 TIMS, ICPMS

α-spec, β-spec, γ-

GKN II 

(Germany) 

SCK•CEN 

(Belgium) 

REBUS 18 × 18 3.8 1 TIMS, ICPM

α-spec, γ-spe

Vandellós II 

(Spain) 

Studsvik Nuclear AB 

(Sweden) 

CSN/ENUSA/ 

ENRESA 

17 × 17 4.498 6 ICPMS, γ-spe

Gösgen 

(Switzerland) 

SCK•CEN, CEA, PSI 

(Belgium, France, 

Switzerland) 

MALIBU 15 × 15 4.3 3 TIMS, ICPMS

α-spec, γ-spe

aKey to laboratory and experimental program abbreviations: bKey to measurement method abbrevATM = Approved Testing Material Program ICPMS = inductively coupled plasm

JAERI = Japan Atomic Energy Research Institute TIMS = thermal ionization mass speYMP = U.S. Department of Energy Yucca Mountain Project LA = luminescence analysis

REBUS, ARIANE, and MALIBU are international programs coordinated by Belgonucleaire, Belgium IDMS = isotopic dilution mass spectCSN = Spanish Regulatory Authority, Consejo de Seguridad Nuclear α-spec = alpha spectrometryENUSA = Spanish fuel vendor, Empresa Nacional del Uranio, S.A. β-spec = beta spectrometryPNNL = Pacific Northwest National Laboratory γ-spec = gamma-ray spectrometry

KRI = Khlopin Radium Institute GE-VNC = General Electric Vallecitos Nuclear Center

ANL = Argonne National Laboratory SCK•CEN = Studiecentrum voor Kernenergie - Centre d'Étude de l'Énergie Nucléaire PSI = Paul Scherrer Institute 

CEA = Commissariat à l'Énergie Atomique ITU = Institute for Transuranium Elements 

3
4 
5 

2 OVERVIEW OF EXPERIMENTAL PROGRAMS AND 

FUEL CHARACTERISTICS 

This section of the report provides a brief description of the experimental programs and the fuel samples 

analyzed. All of the data evaluated in this report were obtained from PWR spent fuel samples. Because 

the NRC research project under which this work was performed focused on analyses for high burnup 

fuels, most of the data are acquired from relatively recent experimental programs that include 

measurements for modern fuel designs and operating characteristics. In addition, the newer programs 

generally include measurements for isotopes of importance to a broader range of spent fuel applications. 

However, several samples from older programs are also included to provide additional data for low and 

moderate burnup fuels such that sufficient data are available to assess trends in the bias and uncertainty. 

Data were acquired from domestic and international programs and include isotopic measurements for 

51 UO2 fuel samples from assemblies irradiated in six different PWRs: 

1. Calvert Cliffs Unit 1 reactor, United States; 

2. Gemeinschaftskernkraftwerk Neckar (GKN) Unit II reactor, Germany; 

3. Gösgen reactor, Switzerland; 

4. Takahama Unit 3 reactor, Japan; 

5. Three Mile Island Unit 1 (TMI-1) reactor, United States; and 

6. Vandellós II reactor, Spain. 

The matrix of spent fuel samples analyzed in this work is listed Table 2.1. Information on the specific 

nuclides measured in each sample and the measurement accuracies can be found in Section 5 of this 

report. 

2.1 Calvert Cliffs 1 

Isotopic measurements for spent fuel samples from the Calvert Cliffs Unit 1 reactor were performed in 

1987 and 1992 at the Material Characterization Center at Pacific Northwest National Laboratory (PNNL) 

for the Approved Testing Material (ATM) Program.6,7

 The measurements were designed to characterize 

mid-burnup spent fuel representative of reactors operating in the United States. Three samples, 

designated ATM-104, were measured from fuel rod MKP109 of a Combustion Engineering (CE) 14 × 14 

assembly identified as D047. Assembly D047 was present in the reactor core for four consecutive cycles, 

from cycle 2 to cycle 5. The assembly had 176 fuel rods and 5 large guide tubes. There were no burnable 

poison rods or gadolinium-bearing rods in the assembly during any of the irradiation cycles. The 

assembly configuration and location of measured rod MKP109 in the assembly are shown in Figure 2.1. 

The samples, identified as 104-MKP109-LL, 104-MKP109-CC, and 104-MKP109-P, had burnup values 

of approximately 27.4, 37.1, and 44.3 GWd/MTU, respectively. The burnup of the samples was 

determined by the laboratory from the 148Nd measurements using the ASTM Standard Test Method E321 

(Neodymium-148 method).8

 The initial enrichment of the MKP109 fuel rod was 3.038 wt % 235U. The 

Calvert Cliffs ATM-104 samples were selected for the present study because extensive data for fission 

products important to burnup credit were measured in these samples. 
6 

Table 2.1. Summary of evaluated spent fuel samples 

Reactor and 

Assembly type Program 

Measurement 

laboratory 

Assembly

ID Fuel type 

Fuel rod 

ID Sample ID 

Enrichment 

(wt % 235U) 

Sample burnup

(GWd/MTU) 

Calvert Cliffs 1 ATM-104 PNNL, KRI D047 UO2 MKP109 MKP109-LL 3.038 27.4 

14 × 14 D047 UO2 MKP109 MKP109-CC 3.038 37.1 

 D047 UO2 MKP109 MKP109-P 3.038 44.3 

Takahama 3 JAERI JAERI NT3G23 UO2 A17 SF95-1 4.11 8.6 

17 × 17 NT3G23 UO2 A17 SF95-2 4.11 14.3 

 NT3G23 UO2 A17 SF95-3 4.11 17.4 

 NT3G23 UO2 A17 SF95-4 4.11 17.7 

 NT3G23 UO2 A17 SF95-5 4.11 25.4 

 NT3G23 UO2-Gd2O3 C13 SF96-1 2.63 24.4 

 NT3G23 UO2-Gd2O3 C13 SF96-2 2.63 29.6 

 NT3G23 UO2-Gd2O3 C13 SF96-3 2.63 30.4 

 NT3G23 UO2-Gd2O3 C13 SF96-4 2.63 30.4 

 NT3G24 UO2-Gd2O3 C13 SF96-5 2.63 30.7 

 NT3G24 UO2 J17 SF97-1 4.11 35.4 

 NT3G24 UO2 J17 SF97-2 4.11 36.7 

 NT3G24 UO2 J17 SF97-3 4.11 40.8 

 NT3G24 UO2 J17 SF97-4 4.11 42.2 

 NT3G24 UO2 J17 SF97-5 4.11 47.0 

 NT3G24 UO2 J17 SF97-6 4.11 47.3 

TMI 1 YMP GE-VNC NJ070G UO2 O1 O1-S1 4.657 25.8 

15 × 15 NJ070G UO2 O1 O1-S2 4.657 29.9 

 NJ070G UO2 O1 O1-S3 4.657 26.7 

 NJ070G UO2 O12 O12-S4 4.657 23.7 

 NJ070G UO2 O12 O12-S5 4.657 26.5 

 NJ070G UO2 O12 O12-S6 4.657 24.0 

 NJ070G UO2 O13 O13-S7 4.657 22.8 

 NJ070G UO2 O13 O13-S8 4.657 26.3 

 ANL NY05YU UO2 H6 A1B 4.013 44.8 

 NY05YU UO2 H6 D2 4.013 44.8 

 NY05YU UO2 H6 B2 4.013 50.1 

 NY05YU UO2 H6 C1 4.013 50.2 

 NY05YU UO2 H6 D1A4 4.013 50.4 

 NY05YU UO2 H6 A2 4.013 50.6 

 NY05YU UO2 H6 C3 4.013 51.3 

 NY05YU UO2 H6 C2B 4.013 52.6 

 NY05YU UO2 H6 B3J 4.013 53.0 

 NY05YU UO2 H6 B1B 4.013 54.5 

 NY05YU UO2 H6 D1A2 4.013 55.7 

GKN II 18 × 18 REBUS SCK•CEN 416 UO2 M11 MA11 3.8 54.1 

Gösgen ARIANE ITU 1601/1701 UO2 P7/R11 GU4 4.1 31.1 

15 × 15 SCK•CEN, ITU 1601/1701 UO2 P7/R11 GU3 4.1 52.5 

 SCK•CEN 12-40 UO2 M13 GU1 3.5 60.7 

 MALIBU SCK•CEN 19-64 UO2 F12 GGU2/2 4.3 46.0 

 PSI 19-64 UO2 F12 GGU2/1 4.3 50.8 

 CEA, PSI 19-64 UO2 F12 GGU1 4.3 70.3 

Vandellos II ENUSA/ Studsvik EC45/EF05 UO2 WZR0058 E58-88 4.498 42.5 

17 × 17 CSN EC45/EF05 UO2 WZR0058 E58-148 4.498 54.8 

 EC45/EF05 UO2 WZR0058 E58-260 4.498 64.6 

EC45/EF05 UO2 WZR0058 E58-700 4.498 77.0 

EC45/EF05 UO2 WZtR165 165-2a 4.498 78.3 

 EC46/EF05 UO2 WZtR160 160-800 4.498 70.9 
7 

 

The ATM-104 data served as the basis for the Organisation for Economic Co-operation and Development 

(OECD)/Nuclear Energy Agency (NEA) burnup credit criticality safety calculation benchmark 

Phase I-B.9

 The assembly design, operating data, and isotopic results are available publicly through the 

NEA Spent Fuel Isotopic Composition Database (SFCOMPO).10 

A B C D E F G H I J K L M N 

1 

2 

3 

4 

5 

6 

7 

8 

9 

10 

11 

12 

13 

14 

Fuel rod

Guide tube 

MKP109 fuel rod

Figure 2.1. Layout for Calvert Cliffs 1 assembly D047. 

2.2 Takahama 3 

The Japan Atomic Energy Research Institute (JAERI) carried out a series of projects from 1990 to 1999 

to obtain high-quality experimental isotopic assay data to support the development of burnup credit for 

storage and transport of spent fuel. The measurements included destructive radiochemical analyses of 

16 spent fuel samples from 3 fuel rods identified as SF95, SF96, and SF97.11,12 Rods SF95 and SF97 

were standard UO2 fuel rods with 4.11 wt % 235U initial enrichment; SF96 was a UO2-Gd2O3 poison rod 

with an initial fuel enrichment of 2.63 wt % 235U and a Gd2O3 content of 6%. Rods SF95 and SF96 were 
8 

from assembly NT3G23, and rod SF97 was from assembly NT3G24. These two assemblies were 

standard 17 × 17 design, with 264 fuel rods and 25 water-filled guide tubes. Fourteen of the fuel rods in 

each of these assemblies contained Gd2O3 poison. 

Assembly NT3G23 was irradiated in the reactor for two consecutive cycles, cycle 5 and cycle 6. 

Assembly NT3G24 was irradiated for three consecutive cycles, starting from cycle 5. The configuration 

of the assembly, including the location of the measured rods, is illustrated in Figure 2.2. The design and 

operating data and isotopic results are available publicly through the Spent Fuel Isotopic Composition 

Database (SFCOMPO) database.15

A B C D E F G H I J K L M N O P Q

1 

2 

3 

4 

5 

6 

7 

8 

9 

10 

11 

12 

13 SF96 

14 

15 

16 

17 SF95 SF97

Guide tube 

Fuel rod

Measured rods 

Gd2O3 rod

Figure 2.2. Layout for Takahama-3 assemblies NT3G23 and NT3G24 and sample locations. 
9 

The sample burnup was determined by JAERI from measurements using the ASTM Standard Test 

Method E 321 (Neodymium-148 method). The burnup of samples from the two UO2 fuel rods, SF95 and 

SF97, were between 14 and 47 GWd/MTU. The burnup of samples from the UO2-Gd2O3 poison rod 

SF96, estimated by JAERI, were between 7.8 and 28.9 GWd/MTU. It was found in a previous study2

 that 

the use of sample burnup values reported by JAERI for the SF96 samples resulted in predicted 148Nd 

concentrations that were 5 to 9% less than the measured concentrations. The differences are likely due to 

a limitation of the ASTM 148Nd method to estimate burnup for fuels containing burnable poisons, since 

the standard method does not account for the energy produced by neutron capture in the gadolinium 

poison. This energy contribution is typically included in most physics codes. Therefore, in this study the 

calculations were performed to a burnup that matched the measured 148Nd content of the fuel. 

2.3 Three Mile Island 1 

Measurements on 19 spent fuel samples from the TMI-1 reactor were performed under the auspices of the 

U.S. Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) Yucca 

Mountain Project in 1998–2000.13 Fuel rods were obtained from two assemblies, identified as NJ05YU 

and NJ070G. Both assemblies are B&W 15 × 15 Mark II design, with 208 fuel rods, 16 guide tubes, and 

1 instrument tube. Radiochemical analyses were performed at two independent analytical laboratories— 

Argonne National Laboratory (ANL) and the General Electric Vallecitos Nuclear Center (GE-VNC). 

Measurements for 11 samples from rod H6 of assembly NJ05YU were performed in 1998 and 2000 at 

ANL.14,15 Fuel rod H6 was irradiated for two cycles (cycles 9 and 10), achieving sample burnups from 

45 to 56 GWd/MTU. Assembly NJ05YU contained 16 burnable poison rods (BPRs) with Al2O3–B4C 

absorber, which were present during cycle 9. All the fuel rods in this assembly had an initial fuel 

enrichment of 4.013 wt % 235U. 

Another eight measured samples were from three rods—O1, O12, and O13—of assembly NJ070G and 

were analyzed in 1999 at GE-VNC.16 Assembly NJ070G was present in the reactor during cycle 10 only, 

and it also contained 16 BPRs during this cycle. Four fuel rods in assembly NJ070G had 2.0 wt % Gd2O3

poison with an initial enrichment of 4.19 wt % 235U. The other 204 regular fuel rods had an initial 

enrichment of 4.657 wt % 235U. The measured samples from rods O1, O12, and O13 achieved burnups 

between 23 and 30 GWd/MTU. The location of the measured rods in both assemblies is shown in 

Figure 2.3. 

Both assemblies were standard commercial designs. However, during cycle 10 the reactor experienced 

anomalies that resulted in significant deposits (crud and/or oxide layers) on fuel rods and fuel failures.17 

The rods used in the present study were selected for postirradiation examinations to investigate the cause 

of the anomalies. Rods O2, O4 through O8, and O11, not included in this study, were known to have 

failed. Consequently, the TMI-1 fuel may have experienced nontypical operating conditions and possible 

geometry perturbations (e.g., rod bowing) and/or temperature variances caused by deviation from normal 

design parameters. The uncertainties in the actual operating conditions need to be considered when 

evaluating calculated TMI-1 results. 
10 

A B C D E F G H I J K L M N O 

1 O1 

2 

3 

4 

5 

6 H6 

7 

8 

9 

10 

11 

12 O12 

13 O13 

14 

15 

 Guide tube 

 Fuel rod 

 Measured fuel rods 

 Gd2O3 rod (in assembly NJ070G only) 

Figure 2.3. Layout for TMI-1 assembly NJ05YU (rod H6) and NJ070G (rods O1, O12, O13). 
11 

2.4 Gösgen 

A total of six UO2 samples were measured from fuel rods irradiated in the Gösgen PWR reactor operated 

in Switzerland. The Gösgen reactor operates with a 15 × 15 fuel assembly design with 205 fuel rods and 

20 guide tubes. Three samples were measured as part of the ARIANE (Actinides Research In A Nuclear 

Element) international program18,19,20 managed by Belgonucleaire. Three additional high burnup samples 

were analyzed in the MALIBU (Radiochemical Analysis of MOX and UOX LWR Fuels Irradiated to 

High Burnup) program,21 now managed by Studiecentrum voor Kernenergie–Centre d'Étude de l'Énergie 

Nucléaire (SCK•CEN) in Belgium. Measurements have been carried out on both low enriched uranium 

oxide (UOX) and mixed oxide (MOX) fuels under these programs; however, only the UO2 samples are 

evaluated in this report. 

These internationally sponsored programs are designed to provide very high-quality evaluated isotopic 

measurement data for computer code validation. An important element of these programs is the use of 

independent laboratories to cross-check radiochemical assay measurements to reduce the measurement 

uncertainties and improve confidence in the data. The laboratories participating in the ARIANE program 

were SCK•CEN in Belgium, Paul Scherrer Institute (PSI) in Switzerland, and the Institute for 

Transuranium Elements (ITU) in Germany. The MALIBU program laboratories included SCK•CEN, 

PSI, and Commissariat à l'Énergie Atomique (CEA) in France. 

ARIANE samples evaluated in this study are designated as GU1, GU3, and GU4. Sample GU1 had an 

initial enrichment of 3.5 wt % 235U and was from assembly 12-40, irradiated for four consecutive cycles, 

starting in cycle 12. The GU1 fuel rod resided next to a control rod guide tube. Several of the fuel rods 

adjacent to the GU1 rod were replaced in the last two irradiation cycles (14 and 15). The other two 

samples, GU3 and GU4, were obtained from a rod of assembly 1601 with an initial enrichment of 

4.1 wt % 235U and irradiated for three consecutive cycles, starting in cycle 16. The GU3/4 rod was 

irradiated in assembly 1601 during cycle 16 and cycle 17 and in assembly 1701 during cycle 18. 

Assembly 1701 had an initial enrichment of 4.3 wt % 235U and was previously irradiated during cycle 17. 

The GU1, GU3, and GU4 samples had burnups of 60.7, 52.5, and 31.1 GWd/MTU, respectively. 

The MALIBU samples, designated GGU1, GGU2/1, and GGU2/2, were obtained from the same fuel rod 

of assembly 19-64. The initial enrichment of the fuel in this assembly was 4.3 wt % 235U. Sample GGU1 

was taken from the location of peak axial burnup of 70.3 GWd/MTU, and samples GGU2/1 and GGU2/2 

were obtained from the lower end of the rod and had a burnup of 50.8 and 46.0 GWd/MTU, respectively. 

Assembly 19-64 was irradiated four consecutive cycles, from cycle 19 to cycle 22. The layout of the 

Gösgen assemblies showing the location of the measured rods is illustrated in Figure 2.4. 

The Gösgen reactor started operating with MOX fuel assemblies in 1997. Therefore, the UO2 fuels 

measured under the ARIANE program (discharged in 1994) were not exposed to MOX fuel. However, 

the UO2 fuel measured under the MALIBU program was exposed to MOX fuel. Typically, MOX cores 

operate with about 1/3 MOX fuel assemblies, while the other assemblies are standard UO2 assemblies. 

The impact of MOX fuel residing adjacent to the measured assemblies was investigated and found to be 

minor. Moreover, the GGU2 fuel rod was located away from the periphery of the assembly and was not 

subject to spectral and flux gradient effects due to the presence of the MOX fuel. 

The relocation of the measured fuel rods in the ARIANE program adds considerable complexity and 

some additional uncertainty to the analysis. For the GU1 rod, the replacement rods in assembly 12-40 

were indicated to have a burnup very similar to that of the original rods. Therefore, explicit modeling of 

the rod replacement was not considered to be necessary. However, for the GU3/4 rod, the second 

assembly (1701) had a burnup considerably lower than the original assembly (1601). To prevent 
12 

potentially biasing the calculated results due to the possible spectral changes caused my moving the fuel 

rod to another assembly during the last cycle, the calculations explicitly modeled the burnup of fuel rods 

in assemblies 1601 and 1701, and included the reconstitution of fuel rods in assembly 1701 in cycle 18. 

A B C D E F G H K L M N P R S 

1 

2 

3 

4 

5 

6 

7 

8 

9 

10 

11 

12 

13 

14 

15 

Guide tube 

Fuel rod

Measured rod for samples GGU1, GGU2/1, GGU2/2

Measured rod for sample GU1

Measured rod for samples GU3/4 (in assembly 1601)

Measured rod for samples GU3/4 ( in assembly 1701)

Figure 2.4. Assembly layout and positions of Gösgen ARIANE and MALIBU samples. 
13 

2.5 GKN II 

Isotopic measurements were reported in the REBUS22 (Reactivity Tests for a Direct Evaluation of the 

Burnup Credit on Selected irradiated LWR fuel bundles) International Program for one UO2 fuel sample 

from 18 × 18 PWR assembly 416 operated in the GKN II reactor located in Neckarwestheim, Germany. 

The assembly had 300 fuel rods and 24 guide tubes. Twelve of the fuel rods contained Gd2O3 at 7.0 wt % 

and had an initial enrichment of 2.6 wt % 235U; the regular fuel rods had an initial enrichment of 3.8 wt % 

235U. The assembly was irradiated for four consecutive cycles, from cycle 5 to 8, and was discharged in 

1997. The measured sample was obtained from a section of UO2 fuel rod M11 that included 

approximately three fuel pellets. The sample achieved a burnup of 54.0 GWd/MTU. 

The REBUS Program, managed by Belgonucleaire, was dedicated to the validation of computer codes for 

criticality calculations that take into account the reduction of reactivity of spent fuel due to burnup 

(burnup credit). REBUS involved critical measurements in the VENUS facility using spent fuel rod 

segments. These segments were later destructively assayed to measure the isotopic content of the fuel. 

The isotopic results were measured by the SCK•CEN laboratory. 

Although the GKN II reactor currently operates with a MOX core, assembly 416 was irradiated in the 

reactor during a period when it operated only with UO2 fuel and, therefore, was not subject to possible 

effects from adjacent MOX assemblies. Rod M11 was located towards the interior of assembly 416, 

away from the guide tubes and, therefore, is expected to have resided in an asymptotic region of uniform 

flux (Figure 2.5). 

2.6 Vandellós II 

Six spent fuel samples were measured from three fuel rods, identified as WZR0058, WZtR165, and 

WZtR160, from two 17 × 17 fuel assemblies irradiated in the Vandellós II PWR reactor operated in 

Spain. The measurements were performed as part of a high burnup fuel program coordinated by the 

Spanish safety council Consejo de Seguridad Nuclear (CSN), fuel vendor Empresa Nacional del Uranio, 

S.A. (ENUSA), and research organization Empresa Nacional de Residuo Radioactivo (ENRESA).23,24 The 

Vandellós fuel samples from the Spanish program represent the highest burnup samples evaluated in this 

report. The samples had a 4.498 wt % 235U initial enrichment and achieved burnup values from 43 to 

78 GWd/MTU. The isotopic measurements were performed at Studsvik Nuclear Laboratory in Sweden. 

The rods were irradiated for five consecutive cycles, from cycle 7 to cycle 11. From cycle 7 to 10, rods 

WZR0058 and WZtR165 resided in assembly EC45 and rod WZtR160 resided in assembly EC46. The 

initial enrichment of assemblies EC45 and EC46 was 4.498 wt % 235U. After cycle 10, the rods were 

removed from their original assemblies and inserted into rebuilt assembly EF05 and irradiated during 

cycle 11. Assembly EF05, with 4.240 wt % 235U initial enrichment, had a burnup of 26.5 GWd/MTU at 

end-of-cycle (EOC) 10. 

The 17 × 17 assemblies have 264 fuel rods, 24 guide tubes, and 1 instrument tube; no burnable poison 

rods or gadolinium-bearing rods were present in the assembly during any of the irradiation cycles. The 

layout of the assemblies showing the locations of the measured rods in the original and rebuilt assemblies 

is illustrated in Figure 2.6. During cycles 7 to 10, all three rods are located at the periphery of the fuel 

assembly. Assemblies EC45 and EC46 were located at the edge of the core during cycle 10. 
14 

A B C D E F G H K L M N P R S T U V 

1 

2 

3 

4 

5 

6 

7 

8 

9 

10 

11 

12 

13 

14 

15 

16 

17 

18 

Guide tube 

Fuel rod 

Gd2O3 rod 

Measured rod M11 

Figure 2.5. Assembly layout for GKN II (REBUS) sample rod M11. 
15 

Similar to the analysis of the ARIANE GU3/4 samples, modeling the relocation of the Vandellós fuel 

rods in cycle 11 adds considerable complexity to the calculations, but was considered necessary due to the 

significant nature of the reconfiguration. In addition, the location of the assemblies at the core edge in 

cycle 10 resulted in some measured rods being aligned at the core periphery, and subjected to additional 

moderation. To minimize any bias in the calculated results due to approximations in the actual 

configuration, assembly boundary regions were explicitly modeled. The added complexity also adds 

uncertainty in the models and this should be considered when analyzing the validation results. 

Q P O N M L K J I H G F E D C B A 

1 

 2 

3 

4 

5 

6 

7 

8 

9 

10 

11 

12 

13 

14 

15 

16 

17 

Figure 2.6. Assembly layout for Vandellós II samples. 

Standard fuel rod 

Measured rod WZR0058 in cycles 7-10 (A12 of assembly EC45) 

Measured rod WZR0058 in cycle 11 (O5 of assembly EF05) 

Measured rod WZtR165 in cycles 7-10 (Q10 of assembly EC45) 

Measured rod WZtR165 in cycle 11 (M3 of assembly EF05) 

Measured rod WZtR160 in cycles 7-10 (A10 of assembly EC46) 

Measured rod WZtR160 in cycle 11 (E15 of assembly EF05) 

Guide tube 
16 
17 

3 MEASUREMENT METHODS AND UNCERTAINTIES 

The accuracy and precision of the measurements are considerations in the evaluation of experimental data 

and in establishing estimates of bias and uncertainty in computational models. The precision is dependent 

on many factors including the fuel dissolution procedures, isotopic concentration in the sample, chemical 

separations techniques, mass spectrometry analysis methods and instruments, calibration methods, and 

standards. 

Thermal Ionization Mass Spectrometry (TIMS) is generally used for high precision measurements of 

uranium, plutonium, and neodymium isotopes. However, several laboratories use Inductively Coupled 

Plasma Mass Spectrometry (ICPMS), which has lower precision but offers more rapid and cost-effective 

routine analysis of nuclides. These measurements provide the isotopic distribution of mass for each 

element. The measurement of absolute nuclide content requires the use of standard spike solutions 

(isotopic dilution), calibration with external sources, or other techniques such as α or γ counting to 

determine the absolute content for one of the isotopes. Therefore, the accuracy of the measurements is 

highly dependent on both the mass spectrometry methods and the method of calibration. 

This section summarizes the measurement methods and the accuracies as estimated by each laboratory. 

The main experimental methods and the experimental uncertainties, expressed in terms of relative 

standard deviation (RSD), for each isotope measured at each of the laboratories are listed in Table 3.1 and 

Table 3.2. The reported accuracies are, in some cases, estimated based on different measures of 

uncertainty. Some laboratories report the error based only on repeat measurements of standard solutions 

(instrument precision). Other laboratories report errors based on benchmark exercises and past experience 

in analyzing fuel samples that include uncertainties and bias due to sample preparations, chemical 

techniques, instrument precision, and drift. Some cross-check measurements performed by independent 

laboratories provide the most reliable method of confirming the estimated measurement uncertainties. 

Measurement errors were not reported for the PNNL measurements of 133Cs. 

3.1 PNNL Measurements 

The isotopic measurements performed by PNNL under the ATM-104 program at the Materials 

Characterization Center (MCC) were made using the following main analytical methods: 

 γ-spectrometry for 137Cs; 

 α-spectrometry for 241Am and 237Np; 

 β-spectrometry for 99Tc and 90Sr; 

 Isotope dilution mass spectrometry (IDMS) for Nd, U, and Pu nuclides using a calibrated triple 

spike of 150Nd, 233U, and 242Pu; 

 Mass spectrometry (MS) after elemental separation of Cs, for 133Cs and 135Cs; and 

 ICPMS measurements relative to 143Nd and 145Nd for lanthanide elements Sm, Eu, and Gd. 

The lanthanide measurements performed by PNNL were carried out by ICPMS without previous 

chemical separation into individual elements. Therefore, there was isotopic interference for the elements 

corresponding to nuclides with mass numbers 147 (Pm, Sm), 150 (Nd, Sm), 151 (Sm, Eu), and 155 (Eu, 

Gd). The reported data corresponding to these four mass numbers were adjusted by PNNL based on 

calculations in order to infer information for individual isotopes. The PNNL lanthanide data are not 

considered for validation purposes in this report to avoid introducing uncertainties related to the 

calculated adjustments.7

 
Table 3.1. Summary of actinide measurement methods and estimated uncertainties 

Nuclide 

ID 

PNNL KRI JAERI GE-VNC ANL Studsvik SCK•CEN ITU C

Methoda RSDb (%) Method RSD (%) Method RSD (%) Method RSD (%) Method RSD (%) Method RSD (%) Method RSD (%) Method RSD (%) Metho

234U IDMS 1.6 IDMS 1.0 TIMS 0.5 ICPMS 4.4 ICPMS 22.0 TIMS 2.5 TIMS 0.0 TIMS 235U IDMS 1.6 IDMS 0.1 TIMS 0.5 ICPMS 3.7 ICPMS 8.7 TIMS 1.0 TIMS 1.2 TIMS 236U IDMS 1.6 IDMS 2.0 TIMS 0.5 ICPMS 5.8 ICPMS 6.9 TIMS 0.4 TIMS 0.8 TIMS 238U IDMS 1.6 IDMS 0.1 TIMS 0.5 ICPMS 4.2 ICPMS na TIMS 0.3 TIMS 0.0 TIMS 

238Pu IDMS 1.6 IDMS 0.5 α-sp 2.5 α-sp 7.9 ICPMS 7.0 TIMS, α-sp 1.6 TIMS 1.1 TIMS 239Pu IDMS 1.6 IDMS 0.3 TIMS 0.6 ICPMS 5.7 ICPMS 5.8 TIMS 0.3 TIMS 0.3 TIMS 240Pu IDMS 1.6 IDMS 0.3 TIMS 0.6 ICPMS 6.2 ICPMS 6.6 TIMS 0.3 TIMS 0.3 TIMS 241Pu IDMS 1.6 IDMS 0.3 TIMS 0.6 ICPMS 4.6 ICPMS 6.9 TIMS 0.3 TIMS 1.7 TIMS 242Pu IDMS 1.6 IDMS 0.3 TIMS 0.6 ICPMS 6.7 ICPMS 6.9 TIMS 0.3 TIMS 0.3 TIMS 

237Np 

α-sp 1.9 α-sp 10.0 α-sp 2.9 ICPMS 5.6 ICPMS 8.5 ICPMS 10.0 ICPMS 4.8 ICPM

241Am 

α-sp 4.9 MS, α-sp 2.0 TIMS, 

α-sp 3.5 γ-sp 7.1 ICPMS 4.9 TIMS 1.8 ICPMS 5.9 TIMS 

242mAm MS, α-sp 10.0 TIMS, 

α-sp 3.5 ICPMS 3.1 TIMS 5.5 TIMS 

243Am MS, α-sp 5.0 TIMS, 

α-sp 3.5 ICPMS 5.9 ICPMS 8.4 TIMS 1.8 ICPMS 6.7 TIMS 

242Cm MS, α-sp 10.0 TIMS, 

α-sp 10.0 α-sp 16.0 TIMS 

243Cm MS, α-sp 2.0 TIMS, 

α-sp 2.75 γ-sp 36.8 TIMS 

244Cm MS, α-sp 2.0 TIMS, 

α-sp 2.75 ICPMS 10.0 α-sp 1.8 ICPMS 6.4 TIMS 

245Cm MS, α-sp 2.0 TIMS, 

α-sp 2.75 TIMS 2.9 ICPMS 10.2 TIMS 

246Cm MS, α-sp 5.0 ICPMS 22.0 TIMS 10.1 TIMS 247Cm MS, α-sp 10.0 

a Main analytical technique is listed; some nuclides required multiple techniques to eliminate interferences.. 

b Laboratory relative standard deviation (RSD) values shown are the maximum of the measurement errors reported for the samples evaluated in this report. 

18
Table 3.2. Summary of fission product measurement methods and uncertainties 

PNNL KRI JAERI GE-VNC ANL Studsvik SCK•CEN ITU CE

Nuclide 

ID Methoda RSDb (%) Method RSD (%) Method RSD (%) Method RSD (%) Method RSD (%) Method RSD (%) Method RSD (%) Method RSD (%) Metho

133Cs IDMS N/A ICPMS 8.0 TIMS 2.5 ICPMS 1.6 TIMS 134Cs γ-sp 3.0 γ-sp 1.75 γ-sp 14.7 TIMS 2.5 ICPMS 4.1 TIMS 135Cs IDMS 14.0 ICPMS 8.5 TIMS 2.5 ICPMS 1.7 TIMS 137Cs γ-sp 3.5 γ-sp 3.0 γ-sp 1.75 γ-sp 4.8 γ-sp 6.4 γ-sp 2.5 ICPMS 1.5 TIMS 

140Ce ICPMS 2.4 142Ce ICPMS 2.6 144Ce γ-sp 10.0 γ-sp 14.1 γ-sp 5.0 γ-sp 

142Nd IDMS 0.1 ICPMS 7.9 TIMS 5.0 ICPMS 5.1 TIMS 143Nd IDMS 1.0 IDMS 0.1 TIMS 0.75 ICPMS 5.5 ICPMS 4.3 TIMS 0.3 ICPMS 6.2 TIMS 144Nd IDMS N/A IDMS 0.1 ICPMS 4.8 TIMS 0.3 ICPMS 5.9 TIMS 145Nd IDMS 1.0 IDMS 0.1 TIMS 0.75 ICPMS 6.2 ICPMS 4.4 TIMS 0.3 ICPMS 5.9 TIMS 146Nd IDMS N/A IDMS 0.1 TIMS 0.75 ICPMS 4.3 TIMS 0.3 ICPMS 7.4 TIMS 148Nd IDMS N/A IDMS 0.1 TIMS 0.75 ICPMS 7.1 ICPMS 4.5 TIMS 0.3 ICPMS 6.7 TIMS 150Nd IDMS N/A IDMS 0.1 TIMS 0.75 ICPMS 5.2 TIMS 0.3 ICPMS 6.8 TIMS 

147Pm β-sp 9.0 ICPMS 6.8 ICPMS 147Sm MS, LA 3.8 IDMS 0.1 TIMS 0.85 ICPMS 10.1 ICPMS 4.7 TIMS 5.0 ICPMS 10.6 ICPMS 148Sm MS, LA 3.8 IDMS 0.1 ICPMS 4.6 TIMS 0.4 ICPMS 4.0 ICPMS 149Sm MS, LA 20.1 IDMS 0.1 TIMS 0.9 ICPMS 8.1 ICPMS 8.3 TIMS 1.1 ICPMS 21.4 ICPMS 150Sm MS, LA 4.6 IDMS 0.1 TIMS 0.85 ICPMS 5.0 ICPMS 4.5 TIMS 0.4 ICPMS 3.4 ICPMS 151Sm MS, LA 38.5 IDMS 0.1 TIMS 0.85 ICPMS 7.1 ICPMS 5.4 TIMS 0.4 ICPMS 33.8 ICPMS 152Sm MS, LA 4.3 IDMS 0.1 TIMS 0.85 ICPMS 4.5 ICPMS 4.5 TIMS 0.4 ICPMS 3.2 ICPMS 154Sm MS, LA 6.0 IDMS 0.1 ICPMS 4.7 TIMS 0.4 ICPMS 10.6 ICPMS 

151Eu MS, LA 9.9 TIMS 0.9 ICPMS 12.5 ICPMS 8.7 TIMS 1.0 ICPMS 153Eu MS, LA 100.0 TIMS 0.9 ICPMS 5.2 TIMS 1.7 ICPMS 5.5 ICPMS 154Eu MS, LA, γ-sp 3.4 γ-sp 3.0 ICPMS 6.7 γ-sp 3.0 ICPMS 11.9 ICPMS 155Eu MS, LA, γ-sp 8.8 γ-sp 7.2 6.1 γ-sp 4.9 ICPMS 16.1 ICPMS 

154Gd IDMS 3.9 ICPMS 5.9 155Gd IDMS 4.2 TIMS 1.4 ICPMS 8.0 ICPMS 8.3 TIMS 2.5 ICPMS 6.9 156Gd IDMS 3.9 ICPMS 4.6 158Gd IDMS 3.9 ICPMS 4.8 160Gd IDMS 12.1 ICPMS 19.4 ICPMS 

90Sr β-sp 5.7 β-sp 8.0 ICPMS 0.4 TIMS 

95Mo ICPMS 4.2 ICPMS 5.0 ICPMS 1.1 ICPMS 

99Tc β-sp 3.5 ICPMS 8.0 ICPMS 8.8 ICPMS 0.9 ICPMS 

101Ru ICPMS 5.8 ICPMS 12.2 ICPMS 0.9 ICPMS 106Ru γ-sp 5.0 γ-sp 15.0 γ-sp 14.2 ICPMS 4.1 γ-sp 103Rh ICPMS 3.8 ICPMS 4.9 ICPMS 3.3 ICPMS 

109Ag ICPMS 5.9 ICPMS 9.1 ICPMS 

125Sb γ-sp 10.0 ICPMS 9.4 γ-sp 

a Main analytical technique is listed; some nuclides required multiple techniques to eliminate interferences.. 

b Laboratory relative standard deviation (RSD) values shown are the maximum of the measurement errors reported for the samples evaluated in this report. V

 as N/A. 

19
20 

The metallic fission product 99Tc was not evaluated in undissolved residues. Technetium and other 

metallic elements, most notably Rh, Ru, Pd, and Mo, segregate in irradiated fuels into a metallic, 

particulate phase that is virtually insoluble in nitric acid.25 Significant discrepancies between 99Tc 

measurements and predictions for the ATM-103 program26 fuel were attributed to residual material that 

had not completely dissolved. However, significant discrepancies or indication of potentially undissolved 

material were not noted in the ATM-104 reports and the 99Tc results were therefore used in this study. 

3.2 KRI Measurements 

More extensive lanthanide fission product measurements that used chemical separations were later 

performed at the Khlopin Radium Institute (KRI) in Russia27 in 1995, at the request of PNNL, using 

archived ATM-104 fuel samples from rod MKP109. The KRI samples are referenced in other reports, 

and in this report, as 87-81, 87-72, and 87-63, respectively. In addition to measurements of specific 

nuclides 143Nd, 145Nd, 155Eu, 155Gd, 147Sm, 149Sm 151Sm, and 152Sm, KRI provided results for most isotopes 

of Nd, Eu, Sm, and Gd that could be measured by mass spectrometry. 

The measurement methods included: 

 Chemical separation of rare earth elements and transuranics followed by chemical separation of 

lanthanides elements, 

 IDMS for neodymium and gadolinium isotopes using a spike of 142Nd and 160Gd, 

 Luminescent analysis (LA)—laser induced fluorometry for absolute measurement of europium and 

samarium content in the sample, 

 MS for europium and samarium nuclides to determine relative isotope ratios, and 

 γ-spectrometry for 154Eu and 155Eu. 

The experimental results were reported by KRI as the ratio of nuclide mass to 145Nd mass and as the 

nuclide mass percentage relative to the corresponding element total mass. As absolute concentrations 

were not reported by KRI, the 145Nd values measured by PNNL were used as the basis to renormalize the 

KRI data in units of g/g U in order to provide a consistent comparison with PNNL measurement data. A 

study of the KRI and PNNL relative atomic percent results, relative to 145Nd concentration, showed that 

the neodymium isotopic data measured in all three samples at both laboratories agreed to within one 

standard deviation experimental uncertainty.2

 

The experimental error varies with the method and the nuclide under consideration. The uncertainties 

listed for KRI (Sm, Eu, Gd) include the error propagation due to the 145Nd renormalization. 

3.3 JAERI Measurements 

The JAERI radiochemical analysis laboratory performed measurements for the Takahama-3 fuel samples. 

After sample cutting and dissolution, the elements were separated using an ion exchange separation 

column. The following main experimental techniques were used to determine the nuclide 

concentrations:11
21 

 IDMS 

o major actinides: U, Pu 

o lanthanides: Nd, Sm 

 α-spectrometry plus MS 

o Am, Cm 

 γ-spectrometry 

o

106Ru, 134Cs,137Cs, 144Ce, 154Eu, 125Sb 

 α-spectrometry 

o

237Np 

The reported experimental uncertainties were not specific to each sample but were typical values based on 

previous experience. The reported uncertainties are less than 0.5% for all measured isotopes of Pu, Sm, 

Nd, 235U, and 238U. For minor actinides measured by MS and α-spectrometry, the experimental errors are 

larger, in the 2 to 10% range. The measurement errors for nuclides determined by γ-spectrometry are 3 to 

10%. Based on experience with international programs that have used interlaboratory cross-check 

measurements, the reported uncertainties of <1% for U, Pu, Nd, and Sm appear to be unrealistically small. 

It is likely that these values represent the ability of the MS analytical instrument to reproduce results 

based on repeat measurements of solutions and may not account for all potential sources of bias and 

uncertainty. 

3.4 GE-VNC Measurements 

The GE-VNC laboratory performed isotopic measurements16 for the samples selected from three fuel rods 

(O1, O12, O13) from TMI-1 assembly NJ070G. Most of the 32 nuclides measured at GE-VNC were 

determined by high precision TIMS and some through γ- or α-spectrometry. The GE-VNC 

measurements included most actinides and fission products but did not include the group of metallic 

fission products (e.g., Rh, Ru, Tc, Mo, and Ag) or 133Cs. 

The nuclide concentrations in the samples measured by TIMS were determined from measurements of 

spiked and unspiked samples. The experimental errors were reported as relative uncertainty at a 95% 

confidence level. The percent RSD uncertainty, shown in the table, was estimated as half (1/1.96) of the 

reported uncertainty at a 95% confidence level. The percent RSD uncertainty is 0.6% for all plutonium 

nuclides except 238Pu, 0.5% for 235U, and 0.8% for neodymium isotopes. 

The nuclide concentrations were reported as g/g 238U, where the 238U value is the measured concentration 

in the sample after irradiation. To provide a consistent basis for the experimental data, the results were 

adjusted to a common basis of g/g U, where U is the mass of initial uranium (unirradiated). 

3.5 ANL Measurements 

ANL performed radiochemical isotopic analyses14 for 11 samples cut from fuel rod H6 of TMI-1 

assembly NJ05YU. The fuel segments were cut at GE-VNC but were measured only by ANL. The 

analysis samples were prepared by dissolution of small quantities (approximately 0.1–0.2 g aliquots) of 

crushed fuel powder, homogenized to provide a uniformity of <1.5% RSD, as determined by multiple 

measurements of uranium and plutonium isotopes. Analyses were carried out using a quadrupole ICPMS, 

γ-spectrometry, and α-spectrometry to determine the isotopic concentrations of 31 nuclides. 

The measurements at ANL included all of the important actinides and most fission products considered 

important to burnup credit and other spent nuclear fuel safety applications, with the exception of the 133Cs 

and 134Cs isotopes. Measurements included metallic fission products in the main solution; however, 
22 

undissolved residues were not measured. The results were reported relative to the measured 238U content 

in the sample, as g/g 238U, and converted for this study to units of g/g U using the measurement data. 

Two measures of experimental uncertainty were reported by ANL—a within-sample precision and a bias 

uncertainty. The within-sample precision was estimated by ANL as one standard deviation obtained 

through repeated measurements of a fuel sample. The bias uncertainty was estimated from deviations 

observed in two separate measurement campaigns involving duplicate fuel samples (obtained from fuel 

cut from adjacent segments of the rod) performed at different times. Phase I measurements were 

performed in September 1998, and Phase II measurements were performed in May 2000. Therefore, the 

precision and bias uncertainty estimated from these measurements is expected to provide a reliable 

estimate of the experimental error. 

The total uncertainty is observed to be relatively large: 3.7% for 235U, 4.2% for 238U, and 5–8% for 

plutonium isotopes. A large 7.1% measurement error for 148Nd is potentially problematic for isotopic 

validation purposes because 148Nd is used to experimentally determine the burnup of the fuel sample. Any 

error in the burnup estimate will propagate as error in calculated values for all isotopes. This potential 

error must be considered in any evaluation of the data, as described in Section 5 of this report. 

3.6 Studsvik Nuclear AB Measurements 

Studsvik Nuclear AB performed isotopic analysis of spent fuel samples from the Vandellós II reactor. 

Measurements were carried out in two experimental campaigns: Phase I measurements,28, 29, 30 performed 

in 2002–2003, and Phase 2 measurements,31 performed in 2007. As part of the second experimental 

campaign at Studsvik, new measurements were performed to increase the number of available samples 

and improve the accuracy of the previous measurements. A reanalysis of samples E58-88 and E58-793 

was carried out in the second campaign using solutions available from the initial campaign. Two new 

samples were measured in Phase 2: one sample identified as E58-257, cut from rod WZR0058, and one 

sample identified as 160-800, cut from rod WZtR160. 

The measurements were performed using elemental separation and ICPMS for isotopic analysis. The 

measurements were carried out using three main methods: 

 γ-scanning analysis 

 ICPMS with external calibration 

 ICPMS with isotopic dilution 

The measurement data were reported as weight percentages relative to measured 238U in the sample and 

converted in this study to units relative to initial uranium using measured data. 

The magnitude of the errors for the measured nuclide concentrations varies with the experimental method 

and the concentration. For nuclides analyzed by γ-scanning, the reported errors are in the 12–15% range 

except for 137Cs, which is about 6%. For the isotopes determined by ICPMS and external one-point 

calibration, the measurement errors are in the 8–10% range with the exception of 246Cm, reported with an 

uncertainty of 15–22%. In the case of the actinides and fission products measured using isotope dilution, 

the experimental errors range between 2 and 22%, depending on the nuclide. The errors for 235U are in 

the range 3 to 9%; 236U, between 4 and 7%; and 234U, between 7 and 22%. The plutonium nuclides were 

measured with an uncertainty of 3 to 8%. For the fission products analyzed by isotopic dilution, 

including Ce, Nd, Sm, Eu, and Gd (except 160Gd), the experimental errors were, on average, between 2 

and 5%. 
23 

3.7 SCK•CEN Measurements 

SCK•CEN has performed isotopic measurements for fuel samples from the international experimental 

programs managed by Belgonucleaire.21 The following main experimental techniques20 have been 

applied for measurements performed at SCK•CEN: 

 TIMS 

o major (U, Pu) and minor (Am and 245,246Cm) actinides 

o lanthanides: Nd, Sm, 144Ce, 155Gd, 151Eu, 153Eu 

o Cs vector nuclides: 133–135Cs 

 ICPMS with external calibration 

o metallics: 95Mo, 99Tc, 101Ru, 103Rh, 109Ag, 125Sb 

o

237Np 

 γ-spectrometry 

o

106Ru, 137Cs, 144Ce, 154Eu, 155Eu, 243Cm 

 α-spectrometry 

o

242Cm, 244Cm 

 β-spectrometry 

o 90Sr 

The metallic species (95Mo, 99Tc, 101Ru, 103Rh, 105Pd, 108Pd, and 109Ag) frequently do not dissolve 

completely in nitric acid. Undissolved residues were analyzed separately by SCK•CEN and the results 

added to the concentrations measured in the main solutions. Based on experience in the ARIANE 

program, the fraction of undissolved material varies significantly for different samples and laboratories, 

from several percent to more than 50%, and the fraction was not observed to be highly dependent on the 

burnup of the fuel. 

3.8 ITU Measurements 

The fuel isotopic measurement methods used at ITU were similar to those used at SCK•CEN. The 

following two main experimental techniques20 were used for measurements performed at ITU: 

 TIMS 

o major actinides (U, Pu) 

 ICPMS with ID (isotope dilution) analysis 

o all other measured nuclides 

The measurements made exclusive use of ICPMS methods for the fission product nuclides. The reported 

ITU uncertainties are considerably larger than those reported by SCK•CEN. A cross-check analysis of 

the ITU measurements for sample GU3 with SCK•CEN results for the same sample indicated a 

systematic bias for some of the ITU results relative to the higher precision TIMS methods used by 

SCK•CEN. In particular, discrepancies likely associated with the ITU measurements were noted for Am, 

Nd, and 106Ru in the ARIANE report. The isotopic data for sample GU3 were the recommended data of 

the ARIANE report, generally obtained as a weighted average of the results reported by ITU and 

SCK•CEN. The results for sample GU4, measured only by ITU, were in accordance with the ARIANE 

program evaluation criteria. 

3.9 PSI Measurements 

The measurements performed at PSI in Switzerland employed the following experimental techniques: 
24 

 Multi-collector ICPMS with ID 

o actinides: U, Pu, Am, Cm 

o lanthanides: Nd, Sm, Eu, Gd 

o Cs vector nuclides 

o 90Sr 

 Mono-collector (quadrupole) ICPMS with external calibration 

o

237Np 

o metallics: 95Mo, 99Tc, 101Ru, 103Rh, 109Ag 

 γ-spectrometry 

o

106Ru, 125Sb, 144Ce 

3.10 CEA Measurements 

The following main experimental techniques have been applied for measurements performed at 

CEA in France: 

 TIMS 

o actinides: U, Pu, Am, Cm 

o lanthanides: Nd 

o Cs vector nuclides 

o 90Sr 

 Multi-collector ICPMS with ID 

o lanthanides: Sm, Eu, Gd 

 Mono-collector (quadrupole) ICPMS with external calibration 

o metallics: 95Mo, 99Tc, 101Ru, 103Rh, 109Ag 

o

237Np 

 γ-spectrometry 

o

106Ru, 125Sb, 144Ce 
25 

4 COMPUTATIONAL ANALYSIS METHODS 

Computational analysis of the spent fuel isotopic composition was carried out using the two-dimensional 

(2-D) TRITON/NEWT depletion sequence in version 5.1 of the SCALE computer code system,1

 released 

publicly in 2006. Details of the measurements, operating history data, and the computational models and 

have been reported separately in Refs. 2–4. A brief description of the codes and methods employed in the 

computational analysis is provided in this section. 

4.1 TRITON/NEWT 

The TRITON/NEWT depletion sequence32 couples the 2-D arbitrary polygonal mesh transport code 

NEWT with the point depletion and decay code ORIGEN to perform burnup simulations. At each 

depletion step, the neutron transport flux solution from NEWT is used to generate cross sections for the 

ORIGEN calculation; the isotopic composition data resulting from each isotopic depletion step is 

employed in the subsequent transport calculation to obtain updated cross sections for the next depletion 

step in an iterative manner throughout the irradiation history. 

NEWT is a 2-D discrete ordinates (SN) multigroup transport code that uses an Extended Step 

Characteristics method solver. This method allows cells to be defined in the form of arbitrary polygons 

and has an automatic fine grid generation feature. The SN method in NEWT allows arbitrary-order 

angular scattering (PN approximation) and arbitrary quadrature order. NEWT has a coarse-mesh finitedifference (CMFD) accelerator that uses a low-order solution for homogenized cells in a coarse spatial 

grid to substantially reduce the number of iterations needed for flux and eigenvalue convergence. 

TRITON can simulate the depletion of multiple mixtures and regions in a fuel assembly model. This is a 

powerful feature for nuclide inventory analysis of measured fuel rods, as it allows an accurate 

representation of the local flux distribution and environmental effects on a specific fuel rod in the 

assembly. 

4.1.1 Computational Models 

Calculations were performed using assembly models that included as many details as were available. The 

following subsections discuss the modeling approach and provide examples of the assembly models used 

in this study. The Calvert Cliffs assembly model provides an example of the models used for many 

standard commercial assemblies. Some samples, including several Gösgen samples, were obtained from 

fuel rods that experienced movement, or reconfiguration, in the assembly during irradiation. Other 

assemblies, such as from the Vandellós reactor, were located at the edge of the core during one of the 

cycles and required more detailed modeling of the assembly surroundings. Including these details in the 

models results in a more accurate representation of the neutronic environment in the fuel that influences 

the calculated isotopic compositions. Therefore, to increase the accuracy of the assembly models, all 

available information was used in the simulations. The following sections describe these models as 

examples of the modeling methods and level of detail. A complete description of the models used for each 

assembly can be found in the primary reference reports. 

4.1.1.1 Calvert Cliffs 

An example of the NEWT assembly model is illustrated in Figure 4.1 for Calvert Cliffs CE 14 × 14 

assembly D047. Only half of the assembly was modeled using symmetry in the assembly. The measured 

rod MKP109, shown in the lattice and labeled, was depleted separately with the four nearest neighbor 

rods. All other rods in the assembly were depleted as a single material. This was found adequate to 
26 

calculate the local neutronic environment of measured rod MKP109. A similar modeling approach was 

used for all other samples. 

standard fuel rods 

measured fuel rod 

nearest neighbors of measured fuel rod 

guide tube 

moderator 

Figure 4.1. Illustration of the NEWT model used for Calvert Cliffs assembly D047. 

4.1.1.2 Gösgen 

Gösgen assembly 12-40, from which sample GU1 was acquired, was reconfigured after the first two 

cycles (cycles 12 and 13). Several fuel rods adjacent to the GU1 rod were replaced with fuel rods from 

another assembly. A detailed irradiation history for the replacement rods was not available; however, the 

burnup of the replacement rods was very similar to that of the original rods, and the effect of the rod 

reconstitution was considered to be minor and was therefore not simulated. 

A reconfiguration of Gösgen assembly 1601, from which sample rod GU3/4 was acquired, was more 

significant. The GU3/4 rod was removed from assembly 1601 after two cycles and placed in a low 

burnup assembly 1701 and irradiated for one additional cycle. Because of potential flux perturbations 

caused by the fuel rod reconfiguration, the rod movement was explicitly simulated in the calculation 

models. This was accomplished using two separate models—one for assembly 1601 and the second for 

assembly 1701 with the GU3/4 fuel rod isotopic compositions determined from the assembly 1601 model. 

4.1.1.3 Vandellós 

Measured rods WZR0058, WZtR165, and WZtR160 were irradiated for four consecutive cycles (cycles 7 

to 10) in assemblies EC45, EC45, and EC46, respectively, and one cycle (cycle 11) in the rebuilt 

assembly EF05. At the time of the reconfiguration, assembly EF05 had been previously irradiated for 

only one cycle and therefore had a significantly lower burnup than the original EC45 or EC46 assemblies. 

The fuel rod movement was modeled explicitly.4

 Information was available on the adjacent assemblies 

and was used in the simulation models because the measured fuel rods were located at the assembly 
27 

periphery and subjected to the potential influence of surrounding neighbors. The assembly models used 

to perform calculations for the spent fuel samples in fuel rod WZR0058 are illustrated as follows: (1) 

assembly EC45 during the first three cycles (Figure 4.2), (2) assembly EC45 at the outer edge of the core 

during the fourth cycle (Figure 4.3), and (3) assembly EF05 containing rod WZR0058 during the final 

cycle (Figure 4.4).

rod WZR0058 

assembly EC45 rods 

neighbor rod 

neighbor rod 

neighbor rod 

neighbor rod 

E10 assembly rods adjacent to EC45 

Figure 4.2. TRITON assembly model for Vandellós II samples from rod WZR0058 (cycles 7–9). 
28 

rod WZR0058 

assembly EC45 rods 

neighbor rod 

neighbor rod 

neighbor rod 

neighbor rod 

G15 assembly rods adjacent to EC45 

Figure 4.3. TRITON assembly model for Vandellós II samples from rod WZR0058 (cycle 10). 
29 

rod WZR0058 

assembly EF05 rods 

neighbor rod 

neighbor rod 

neighbor rod 

neighbor rod 

clad 

guide tube 

Figure 4.4. TRITON assembly model for Vandellós II samples from rod WZR0058 (cycle 11). 

4.2 Cross-Section Libraries 

Neutron transport calculations were performed using the SCALE 44-group cross-section library that 

contains 22 thermal-upscatter groups. The 44-group library is collapsed from the ENDF/B-V SCALE 

238-group library using an LWR fuel flux spectrum. 

4.3 Resonance Processing 

The NITAWL module in SCALE, which is based on the Nordheim Integral Treatment, was used in this 

study for self-shielding of the resolved resonance cross sections. The NITAWL module was used in this 

study based on its demonstrated performance for LWR fuel analyses. NITAWL, however, cannot be used 

with the ENDF/B–VI and –VII cross-section libraries, and it does not allow fuel rod subdivision (i.e., fuel 

rods must be treated as a single region with one radial zone). 
30 

4.4 Isotopic Depletion Calculations 

Isotope transmutation and decay calculations were performed with the ORIGEN code. Cross sections 

used in the ORIGEN calculations are generated automatically during the TRITON depletion analysis 

using region- and time-dependent cross sections calculated by NEWT. Burnup-dependent cross sections 

for 232 isotopes defined in the 44-group library are generated based on the transport calculation solution 

(addnux = 3 input option). This procedure ensures that cross sections are updated for many of the key 

isotopes of interest to spent fuel safety applications and their capture and/or decay precursors. 
31 

5 ANALYSIS RESULTS AND DISCUSSION 

The experimental isotopic data considered in this study include measurements of more than 60 isotopes 

obtained from 51 spent fuel samples. The analysis is based on the collection of the isotopic validation 

results reported separately.2–4 The data for Gösgen samples measured under the MALIBU experimental 

program are currently commercial proprietary and can not be made publicly available at this time. 

This section describes the isotopes that are important to several spent fuel safety applications, presents the 

results of the comparisons of isotopic calculations with experiment, and reviews the results. Evaluation 

of the results by measurement laboratory provides information on the accuracy and consistency of the 

analytical techniques. Comparison of results from different laboratories can be used to identify potential 

problems related to experimental methods, specific isotopes, or fuel samples. The results are also 

evaluated as a function of sample burnup to identify trends in the bias, particularly as these trends relate 

to high burnup fuel. Recommendations related to the quality and application of the experimental data sets 

are presented. 

5.1 Important Isotopes in Spent Nuclear Fuel Safety Analyses

Uncertainty in the predicted isotopic composition in spent fuel represents one of the largest sources of 

uncertainty in criticality calculations that take credit for the irradiated fuel composition (burnup credit), 

spent fuel decay heat, neutron and gamma radiation source terms, and nuclide concentrations and 

activities used in radiological safety calculations. The isotopes identified as being of highest importance 

to several safety-related applications are described here. 

Measurements for individual isotopes have direct application to the evaluation of uncertainties in safety 

applications involving irradiated nuclear fuel, particularly in areas where direct integral measurements are 

not practical (e.g., spent fuel reactivity measurement for burnup credit) or in application regimes beyond 

the range for which integral measurements are available or feasible (e.g., decay heat at long cooling times 

of importance to repository applications). These measured data may be used to assess uncertainties in 

integral quantities and also identify the sources of bias in integral calculations. Because the isotopic 

measurements are generally performed on fuel that has cooled for several years, many of the short-lived 

nuclides are decayed below the detection limit. Therefore, the nuclides of radiological interest having 

available measurements are generally those important at times greater than about one year after discharge. 

Previous studies have investigated the relative importance of individual nuclides in spent fuel 

applications,33 as well as the changes in the importance of these nuclides for those applications involving 

high burnup fuel.34 Table 5.1 lists 40 nuclides considered important to burnup credit, decay heat and 

radiological source term calculations, and high-level waste management. Measurement data are available 

for all isotopes in this list. 

The actinides and fission products of highest importance to the neutron multiplication factor in spent fuel 

are well documented elsewhere.35 Actinides widely considered for burnup credit for spent fuel storage 

and transport cask criticality safety analyses are 234U, 235U, 238U, 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, and 

241Am. These nine actinides represent about 95% of the reactivity worth of the actinides and about 70% 

of the total reactivity worth of all nuclides in typical spent fuel. Six major fission products, 143Nd, 149Sm, 

103Rh, 151Sm, 133Cs, and 155Gd, account for about 75% of the fission product reactivity worth, and about 

20% of the total reactivity worth. The relative worth of these nuclides will vary to some degree, 

depending on the enrichment, burnup, assembly design, and cooling time, but the list of important 

nuclides will remain the same. 
32 

Table 5.1. Important nuclides in spent fuel safety applications 

Isotope Half life Burnup 

credit 

Radiological

safety 

Waste 

management 

Se-79 2.95 × 105 years ■

Mo-95 Stable ■ 

Sr-90/Y-90 28.9 years ■ ■

Tc-99 2.111 × 105 ■ ■

Ru-101 Stable ■ 

Ru-106 371.6 days ■ 

Rh-103 Stable ■ 

Ag-109 Stable ■

Sb-125 2.7586 years ■ 

I-129 1.6 × 107 years ■

Cs-133 Stable ■ 

Cs-134 2.065 years ■ 

Cs-135 2.3 × 106 years ■

Cs-137/Ba-137a 30.0 years ■ ■

Nd-143 Stable ■ 

Nd-145 Stable ■

Nd-148 a Stable 

Ce-144/Pr-144 284.9 days ■ 

Gd-155 Stable ■ 

Sm-147 1.06 × 1011 years ■ 

Sm-149 Stable ■ 

Sm-150 Stable ■

Sm-151 90 years ■ 

Sm-152 Stable ■ 

Eu-151 Stable ■ 

Eu-153 Stable ■ 

Eu-154 8.59 years ■ 

U-234 2.455 × 105 years ■ ■

U-235 7.037 × 108 years ■ ■

U-236 2.342 × 107 years ■ ■

U-238 4.468 × 109 years ■ ■

Np-237 2.14 × 106 years ■ ■

Pu-238 87.71 years ■ ■ ■

Pu-239 2.41 × 104 years ■ ■ ■

Pu-240 6.56 × 103 years ■ ■ ■

Pu-241 14.29 years ■ ■

Pu-242 3.75 × 105 years ■ ■

Am-241 433 years ■ ■ ■

Am-243 7370 years ■ ■

Cm-242 162.8 days ■ 

Cm-244 18.1 years ■ a

 Isotopes commonly used for experimental burnup determination.
33 

The relative contribution to reactivity of actinides and fission products for typical spent fuel in a highcapacity storage cask after 5 years of cooling is illustrated in Figure 5.1.36 The measure of nuclide 

importance was determined from the relative sensitivity of the neutron multiplication factor to the 

concentration of each individual nuclide in the fuel. The fission product 131Xe is shown in the figure, but 

it is not generally considered in burnup credit because of its volatility. U-235 Pu-239 U-238 Pu-240 Pu-241 Sm-149 Nd-143 Rh-103 Sm-151 Am-241 Xe-131 U-236 Cs-133 Gd-155 Np-237 Sm-152 Tc-99 Nd-145 Eu-153 Pu-242 Sm-147

Mo-95

Pu-238

Sm-150

Ag-109

Ru-101

Pd-105

Pr-141

Am-243

Eu-151

U-234

10-3

10-2

10-1

Relative importance

Figure 5.1. Relative isotopic importance in burnup credit criticality calculations. 

Figure 5.2 shows the relative percent contribution of the most important isotopes to decay heat for cooling 

times up to 1000 years for typical high burnup fuel. These isotopes contribute about 95% to the total 

decay heat after 1 year, and more than 99% to the total decay heat beyond 2 years. Measurements of fulllength assembly decay heat have been reported for cooling times from about 3 to 30 years cooling, and 

these measurements have been used to validate code calculations over this timeframe.37 However, 

measurements for cooling times beyond 30 years have not been made. 

After 30 years, fission product decay heat is dominated by 137Cs and 90Sr, and their decay progeny, 137mBa 

and 90Y, respectively. Actinide decay heat becomes increasingly important at longer cooling times, with 

241Am, 238Pu, and 244Cm being the largest contributors. Isotopic measurements of the dominant nuclides 

have been made and can be applied to establish the accuracy of code calculations beyond the regime 

where integral measurements are available. Although measurements for 137mBa and 90Y are not available, 

these nuclides are in secular equilibrium with the parent nuclides, 137Cs and 90Sr, and can therefore be 

accurately derived from the concentration of the parents. 
34 

Figure 5.2. Relative isotopic contribution to the total decay heat for typical high burnup fuel

0 5 10 15 20 100 100

0.1

1

10

100

144Pr

125Sb

144Ce

241Pu

244Cm

242Cm

106Rh

147Pm 134Cs

154Eu 

137Cs

90Sr 

137mBa 

90Y 

238Pu 2

240

Fraction of total heat (percent)

Decay time (years)

241Am
35 

The isotopes important in radiation shielding applications are similar to those important in decay heat 

calculation. The dose rates from activated fuel components (e.g., 60Co) are an exception. After 10 years, 

the contribution from 137mBa (Eγ = 662 keV), the decay progeny of 137Cs, is by far the largest component 

of the shielded dose rate. Other important nuclides include 244Cm (spontaneous fission), 154Eu, and 90Y 

(Eγ = 2.186 MeV), the decay progeny of 90Sr. For cooling times less than 10 years, the relative 

contributions from 134Cs, 106Ru, and 144Pr (the decay progeny of 144Ce) are dominant. Measurements for 

these nuclides are available and have been analyzed in this report. The relative isotopic contributions to 

the dose rates have been calculated for several cask designs and reported previously.34

Measurement data on the concentration of isotopes in spent fuel are required to support the waste 

management and safety assessments of spent fuel repositories. However, the isotopes of interest for waste 

disposal are different from those important to other applications (e.g., criticality safety). In the field of 

safety assessment of spent fuel and vitrified high-level waste repositories, long-lived fission products are 

very important radionuclides in addition to long-lived actinides. For fission product nuclides, there is 

broad international interest in the spent fuel and high-level vitrified waste management community in 14C, 

36Cl, 79Se, 99Tc, 126Sn, 129I and 135Cs. Measurements available for the long-lived fission products important 

to waste management safety assessments are much more limited, and additional experimental data are 

needed to fully support this application. 

5.2 Comparison of Calculated and Experimental Results 

Tables 5.2–5.7 list the relative percent error between the calculated and experimental nuclide 

concentrations for each analyzed sample. The methods used to measure each isotope and the 

experimental error reported by each laboratory can be found in Section 3. Plots showing the comparisons 

of calculated and measured results are provided in Appendix A (Figures A.1–A.58). For completeness, 

all isotopes with measured data are included in these plots. The results are plotted as a function of sample 

burnup and by the measurement laboratory. The results, grouped by laboratory, can aid in identifying 

correlated bias associated with measurement methods. The error bars shown in the plots were calculated 

by propagating the reported standard error of the measured nuclide concentrations to the plotted values of 

percent difference between calculation and experiment. 

The error bars shown in the figures do not account for the additional uncertainty in the sample burnup as 

determined on the basis of measured 148Nd concentration. In general, the measurements of 148Nd were 

made using high-precision methods involving isotopic dilution and thermal ionization mass spectrometry. 

However, some laboratories reported difficulties measuring 148Nd or used lower-precision methods that 

resulted in much larger uncertainties. Error in the sample burnup caused by measurement uncertainty will 

be manifested as code bias. While the burnup error is evident through the calculations, it derives from 

experimental quantities. Error in the sample burnup will affect the calculated concentrations of each 

isotope to a different extent. Including the influence of the burnup uncertainty, if available, in the error 

bars of C/E comparisons would provide a more realistic measure of how experimental uncertainty impacts 

the accuracy of the calculations. 
36 

Table 5.2. C/E-1 (%) for Calvert Cliffs samples 

Sample ID 87-81 

(104-MKP109-LL)

87-72 

(104-MKP109-CC)

87-63 

(104-MKP109-P) 

Burnup

(GWd/MTU) 27.35 37.12 44.34 

U-234 -1.4 -2.7 2.2 

U-235 -1.5 -2.4 -1.1 

U-236 2.1 2.4 1.8 

U-238 -0.7 -0.5 -0.2 

Pu-238 -9.6 -7.3 -6.6 

Pu-239 2.5 3.5 6.2 

Pu-240 0.0 0.1 0.7 

Pu-241 -2.5 -2.2 -0.2 

Pu-242 -0.9 -0.3 -2.1 

Np-237 6.4 15.5 6.9 

Am-241 -4.6 -9.8 -8.1 

Cs-133 0.7 1.9 3.1 

Cs-134 -4.8 -14.8 -20.8 

Cs-135 6.2 5.0 4.6 

Cs-137 -0.8 -0.4 -1.0 

Nd-143 0.5 0.9 2.2 

Nd-144 -0.8 -1.0 -1.2 

Nd-145 -0.6 -0.9 -0.9 

Nd-146 0.7 0.8 0.9 

Nd-148 0.0 0.0 0.0 

Nd-150 2.2 3.2 4.1 

Sm-147 3.8 0.3 -8.6 

Sm-148 -1.1 -1.3 -8.3 

Sm-149 -26.3 28.8 -42.4 

Sm-150 6.4 8.2 4.8 

Sm-151 42.7 30.9 30.3 

Sm-152 23.61 30.8 24.3 

Sm-154 -11.4 8.3 -6.3 

Eu-151 -43.4 23.5 -57.4 

Eu-152 -69.0 -48.4 -93.8 

Eu-153 3.2 3.0 3.4 

Eu-154 -4.2 2.5 9.3 

Eu-155 -31.7 -29.9 -30.1 

Gd-154 -20.7 32.3 32.7 

Gd-155 -48.7 -26.4 -28.8 

Gd-156 -24.9 37.2 64.6 

Gd-158 -19.1 -99.4 -99.4 

Gd-160 -48.6 

Tc-99 5.2 6.8 14.1 

Sr-90 3.2 1.1 2.4 
37 

Table 5.3. C/E-1 (%) for Takahama-3 samples 

Sample ID SF95-1 SF95-2 SF95-3 SF95-4 SF95-5 SF97-1 SF97-2 SF97-3 SF97-4 SF97-5 SF97-6 SF96-1 SF96-2 SF9

Burnup 

(GWd/MTU) 14.3 24.4 35.4 36.7 30.4 17.7 30.7 42.2 47.0 47.3 40.8 8.6 17.4 29.

U-234 9.4 -1.4 26.4 25.1 -8.2 6.4 9.6 6.8 6.6 7.2 8.4 -7.6 -5.9 -7U-235 0.9 2.7 3.1 3.9 2.1 3.3 0.7 1.4 1.9 0.6 2.0 1.9 1.9 3U-236 0.6 -1.7 -0.1 -0.3 -1.0 1.2 -0.2 -0.3 -0.4 -0.3 -0.7 -2.0 -1.6 -0Pu-238 8.9 -3.7 4.9 4.4 2.0 31.7 -5.5 -7.6 -10.4 -12.6 -7.0 20.4 -1.4 -1Pu-239 12.8 8.0 8.7 7.9 8.3 31.7 4.0 4.1 3.3 0.8 4.9 22.0 3.8 1Pu-240 7.0 3.4 5.9 6.2 6.2 14.9 7.0 7.6 6.3 5.8 7.3 13.2 7.4 7Pu-241 11.4 0.6 1.0 1.6 2.3 30.4 -2.6 -2.5 -3.0 -5.4 -1.4 26.4 1.9 1Pu-242 13.6 0.0 -0.8 -0.1 3.1 23.3 1.0 -1.1 -2.4 -2.9 -0.8 21.9 5.6 3Np-237 24.5 1.8 3.8 1.2 -1.9 -0.6 36.5 45.1 65Am-241 -6.5 23.4 23.9 70.3 18.7 74.5 29.2 27.9 15.4 10.4 31.9 67.6 45.0 29Am-242m 22.5 17.5 17.3 19.7 22.5 111.4 25.8 18.0 6.7 1.7 18.9 31.0 9.0 23Am-243 35.8 22.2 21.7 23.8 23.8 77.0 15.3 13.6 10.9 9.0 15.4 65.4 25.8 24Cm-242 -16.0 -30.5 -37.5 -45.5 -18.1 19.6 -4.4 1.9 6.9 12.6 2.2 0.3 -18.6 -17Cm-243 -6.3 -19.2 -10.5 -13.5 -20.9 29.9 -18.5 -17.4 -17.9 -20.6 -17.8 Cm-244 30.3 -0.3 8.1 5.8 13.0 86.6 -2.3 -4.0 -6.8 -9.3 -1.1 51.8 7.3 6Cm-245 13.9 -21.5 -13.9 -19.7 -14.1 79.9 -28.4 -30.2 -33.1 -36.3 -26.2 Cm-246 -45.1 -66.1 -21.2 -23.8 25.0 48.2 -93.2 -34.0 -38.2 -40.7 -33.5 Cm-247 -31.5 -32.1 -35.9 -38.4 -36.0 Nd-142 -19.0 -7.3 -15.3 -13.1 -5.4 Nd-143 -2.0 -2.0 -1.9 -1.0 -1.4 0.4 0.2 0.2 0.7 -0.2 0.3 -4.2 -2.6 -1Nd-144 0.0 -1.6 -1.0 -4.7 -1.2 3.3 1.4 -2.1 -3.9 -4.3 -2.3 -6.8 -6.0 -10Nd-145 0.3 -0.3 -0.5 -0.1 0.2 0.4 1.3 0.9 1.1 1.1 0.8 -1.2 0.3 0Nd-146 2.8 1.8 1.7 1.6 2.3 2.4 1.5 0.8 0.7 0.6 1.0 -1.0 0.1 0Nd-148 1.0 -0.3 -0.5 -0.7 -0.1 1.7 0.9 0.2 0.0 0.0 0.4 -0.2 0.0 -0Nd-150 -0.2 0.4 -0.4 -0.3 1.1 4.3 1.9 1.6 1.3 1.4 2.1 0.7 0.7 1
38 

Table 5.3 (continued) 

Sample ID SF95-1 SF95-2 SF95-3 SF95-4 SF95-5 SF97-1 SF97-2 SF97-3 SF97-4 SF97-5 SF97-6 SF96-1 SF96-2 SF9

Burnup 

(GWd/MTU) 14.3 24.4 35.4 36.7 30.4 17.7 30.7 42.2 47.0 47.3 40.8 8.6 17.4 29

Cs-134 -8.6 -13.9 -12.5 -12.7 -12.8 -5.4 -19.7 -17.6 -14.9 -15.4 -16.2 1.1 0.9 1Cs-137 -1.6 -3.0 -2.3 -2.7 -1.6 -2.1 -2.3 -2.6 -1.8 -2.0 -2.6 5.2 -8.0 -4Ce-144 -2.1 -2.0 -5.0 4.2 -1.7 -17.7 -10.2 -2.7 6.0 6.7 -4.8 -0.3 5.1 15Eu-154 7.3 0.8 5.1 1.8 3.2 21.4 -2.7 1.3 1.9 0.3 2.4 5.0 -2.3 7Ru-106 2.8 20.5 29.2 30.4 12.5 -6.5 -4.7 -2.2 8.3 80.4 -15.0 27.5 53.0 36Sb-125 94.9 86.3 126.3 175.9 112.8 31.4 22.5 84.4 70.6 38.6 91.3 30.5 52.8 11Sm-147 -2.3 0.6 -1.7 -2.9 -2.3 -0.9 Sm-148 17.8 -4.2 -9.6 -12.7 -13.6 -8.1 Sm-149 3.2 -8.5 -2.5 6.1 6.3 -3.0 Sm-150 5.4 4.9 5.1 5.0 4.4 6.4 Sm-151 51.4 30.0 34.3 33.3 28.0 33.9 Sm-152 7.2 23.1 29.1 30.8 30.3 26.3 Sm-154 3.5 0.2 0.3 -0.2 -1.1 1.0 
39 

Table 5.4. C/E-1 (%) for TMI-1 samples measured at GE-VNC 

Sample ID O13S7 O12S4 O12S6 O1S1 O13S8 O12S5 O1S3 O1S2 

Burnup 

(GWd/MTU) 22.8 23.7 24.0 25.8 26.3 26.5 26.7 29.9 

U-234 -2.4 -0.1 0.5 -0.4 0.6 2.3 0.9 0.5 

U-235 3.8 3.6 1.3 3.7 4.0 3.9 2.4 5.3 

U-236 -3.0 -3.9 -4.0 -3.5 -3.0 -3.4 -2.9 -2.6 

Pu-238 -15.2 -15.5 -22.9 -12.7 -18.6 -18.5 -18.9 -11.7 

Pu-239 2.2 2.3 -4.1 3.0 2.0 0.0 0.4 9.8 

Pu-240 -0.9 -0.8 -4.1 -0.8 -2.4 -3.6 -4.8 -1.3 

Pu-241 -6.8 -6.1 -11.8 -5.4 -8.3 -9.6 -9.6 -2.9 

Pu-242 -10.7 -9.8 -11.8 -8.5 -11.8 -12.2 -13.3 -11.8 

Np-237 -2.7 -7.6 -7.8 1.7 -3.8 -3.4 -4.7 0.5 

Am-241 -4.6 -9.0 9.7 33.2 -6.3 -8.3 0.7 12.2 

Am-242m 4.4 -4.9 2.2 35.2 -10.2 -13.1 3.7 15.7 

Am-243 1.1 0.9 20.9 50.1 -1.7 -3.6 13.0 19.1 

Cm-242 -24.2 -28.9 -21.4 -10.7 -37.5 -34.2 -30.3 -40.0 

Cm-243 -29.1 -29.2 -22.5 6.8 -31.4 -34.2 -25.3 -13.5 

Cm-244 -15.9 -16.7 -6.2 30.1 -19.4 -21.6 -7.0 2.2 

Cm-245 -42.2 -42.7 -42.1 -8.1 -46.4 -47.9 -37.9 -23.3 

Nd-143 2.2 2.0 2.1 2.7 2.6 2.5 2.8 3.4 

Nd-145 1.4 1.2 2.1 1.6 1.7 1.8 2.2 1.6 

Nd-146 0.4 0.4 0.3 0.7 0.3 0.3 0.9 0.9 

Nd-148 0.0 0.0 0.0 -0.1 0.0 0.0 0.0 0.0 

Nd-150 -0.1 0.6 -0.2 0.0 0.2 0.1 -0.3 0.3 

Cs-134 -23.1 -22.4 -23.6 -20.1 -23.4 -22.5 -21.1 -18.2 

Cs-137 -5.9 -4.5 -3.6 -3.4 -6.4 -4.9 -4.1 -7.2 

Sm-147 -1.8 -4.9 -3.7 -5.1 -2.4 -3.0 -5.2 -4.7 

Sm-149 13.4 11.6 8.5 14.6 19.0 18.6 13.6 28.2 

Sm-150 1.7 1.0 1.4 2.0 1.6 1.0 1.4 2.4 

Sm-151 28.8 27.1 18.0 30.1 24.8 24.9 24.8 32.8 

Sm-152 13.8 13.4 17.9 14.7 16.9 16.6 15.4 14.2 

Eu-151 33.3 19.8 11.6 24.3 29.1 28.4 20.5 39.8 

Eu-153 -7.7 -8.8 -8.4 -6.1 -8.5 -8.1 -6.5 -5.1 

Gd-155 -38.7 -42.4 -47.1 -46.2 -41.9 -40.9 -48.2 -36.7 
40 

Table 5.5. C/E-1 (%) for TMI-1 samples from assembly NJ05YU measured at ANL 

Sample ID A1B D2 B2 C1 D1A4 A2 C3 C2B B3J B1B D1A2 

Burnup

(GWd/MTU) 44.8 44.8 50.1 50.2 50.5 50.6 51.3 52.6 53.0 54.5 55.7 

U-234 5.6 10.0 5.6 -0.8 1.0 2.4 4.8 7.6 5.7 1.2 -3.4

U-235 0.9 24.7 14.5 11.0 -2.5 9.0 9.2 5.2 1.7 -6.5 -16.1

U-236 4.9 0.5 2.5 1.2 3.3 1.0 4.5 8.0 2.8 4.5 3.6

Pu-238 -34.6 -12.3 7.4 5.8 -6.1 -3.5 41.9 -18.8 -7.8 -8.8 9.7

Pu-239 14.6 17.2 12.3 14.7 16.5 9.9 9.1 22.3 15.1 17.7 14.0

Pu-240 19.0 7.0 11.1 11.4 18.1 9.2 8.5 23.9 18.0 22.1 21.1

Pu-241 2.0 5.3 5.5 6.6 -0.6 7.3 6.9 7.0 1.1 5.0 1.2

Pu-242 8.6 -7.8 0.6 1.9 -1.8 1.7 4.0 7.8 -7.4 12.4 15.0

Np-237 2.5 -4.3 1.7 2.0 6.2 1.8 6.1 9.0 4.8 9.6 12.5

Am-241 8.2 -14.8 -13.6 -18.7 -18.3 -3.0 -0.5 -15.9 -18.4 48.4 32.4

Am-242m 75.8 -17.1 4.7 31.0 133.0

Am-243 46.6 -1.7 -1.8 3.7 40.7 1.0 8.6 46.6 37.1 53.2 61.4

Nd-143 4.6 14.9 8.8 12.0 2.2 14.3 15.6 7.9 4.2 3.5 2.7

Nd-145 3.4 5.8 5.3 6.2 -0.4 9.5 8.1 4.8 1.6 2.8 2.2

Nd-148 0.8 0.9 0.4 0.5 0.5 0.3 0.3 0.3 0.2 0.3 -0.1

Cs-137 -13.7 -6.7 -3.7 -6.9 -1.3 -3.8 1.3 -3.6 -1.3 0.1 17.0

Sm-147 11.0 17.8 19.2 17.7 7.2 12.8 21.7 11.4 3.5 0.6 1.2

Sm-149 16.9 27.0 18.5 25.8 12.3 0.9 36.3 20.6 23.2 18.9 9.5

Sm-150 8.8 12.1 17.1 15.1 7.6 18.7 24.6 10.7 3.0 3.0 8.7

Sm-151 35.8 54.4 41.7 59.4 41.5 50.1 55.0 48.5 28.6 31.6 33.0

Sm-152 34.5 33.8 37.9 40.2 32.7 36.3 44.3 41.8 31.4 32.5 34.7

Eu-151 1.5 -30.8 -40.8 -28.2 13.7 -47.3 -43.3 6.3 -4.0 31.0 17.9

Eu-153 7.6 1.7 8.2 8.6 4.8 7.1 15.8 11.1 5.3 7.5 8.0

Eu-155 -61.5 -59.2 -55.5 -58.8 -63.2 -53.9 -52.5 -50.7 -52.2 -66.7 -45.9

Gd-155 -52.0 -47.9 -48.4 -46.3 -65.9 -34.5 -47.5 -46.8 -51.8 -47.7 -46.9

Mo-95 -3.9 8.2 -3.2 -0.8 0.5 -1.6 10.4 3.3 1.5 1.4 6.3

Tc-99 -26.0 7.4 5.3 6.3 -3.2 7.1 13.1 -12.1 -3.5 -6.7 9.2

Ru-101 -6.5 9.9 -3.6 -0.4 6.1 1.2 15.6 3.5 4.3 5.7 13.0

Rh-103 2.6 19.5 5.5 7.8 11.1 7.7 23.2 11.8 10.8 12.1 15.6

Ag-109 100.7 123.4 127.2 125.3 44.0 103.2 34.2 96.5 65.6 205.4 200.1
41 

Table 5.6. C/E-1 (%) for Gösgen and GKN II samples 

Sample ID GU1 GU3 GU4 GGU1 GGU2/1 GGU2/2 M11 

Laboratory SCK•CEN SCK•CEN, 

ITU ITU CEA, PSI 

PSI SCK•CEN SCK•CEN 

Program ARIANE ARIANE ARIANE MALIBU MALIBU MALIBU REBUS 

Burnup 

(GWd/MTU) 60.7 52.5 31.1 70.3 50.8 46.0 54.0 

U-234 15.6 37.4 41.4 -0.7 0.4 3.3 19.7 

U-235 4.4 -1.4 -0.3 -2.5 -6.6 0.2 4.3 

U-236 0.4 -0.4 0.1 -2.1 0.4 1.1 -0.6 

U-238 -0.3 -0.9 -0.5 -1.4 -0.1 1.0 0.0 

Pu-238 -2.4 -5.4 -2.7 -4.4 -10.1 -10.4 -7.6 

Pu-239 5.6 1.3 4.7 5.1 -4.2 -0.3 8.5 

Pu-240 2.9 4.0 3.6 3.0 1.7 3.4 3.3 

Pu-241 -0.6 -4.4 -2.6 -2.2 -6.7 -6.4 0.8 

Pu-242 -3.1 -1.0 0.7 -4.7 0.7 -1.8 -2.1 

Np-237 -9.6 -27.1 3.0 7.4 10.0 27.0 

Am-241 8.0 11.7 -3.2 -8.4 -10.7 13.4 28.9 

Am-242m 49.4 18.2 29.3 -17.2 29.5 27.3 

Am-243 17.5 20.6 22.8 2.8 7.5 30.7 37.5 

Cm-242 -15.5 -22.0 -10.7 -21.2 26.6 

Cm-243 202.3 22.3 -16.0 -6.3 -7.9 

Cm-244 2.1 -7.8 -12.5 -6.4 -1.5 -9.1 -6.1 

Cm-245 -19.9 -41.6 -35.4 -27.5 -29.4 -33.7 -31.6 

Cm-246 -36.7 -38.2 -37.2 -29.0 -34.0 

Nd-142 4.7 5.0 18.1 2.2 4.7 -12.0 -8.1 

Nd-143 6.9 2.1 -2.1 0.9 -2.5 3.4 4.1 

Nd-144 0.3 0.0 -3.3 -4.4 -2.7 1.2 -1.3 

Nd-145 2.3 0.6 -0.8 -3.0 -2.5 2.5 0.9 

Nd-146 3.1 0.2 -0.2 -2.6 -2.1 2.0 1.1 

Nd-148 0.0 -0.4 -0.1 -4.1 -2.9 1.1 0.3 

Nd-150 0.9 -0.2 2.4 -2.1 -0.8 2.0 2.2 
42 

Table 5.6 (continued) 

Sample ID GU1 GU3 GU4 GGU1 GGU2/1 GGU2/2 M11 

Laboratory SCK•CEN SCK•CEN, 

ITU ITU CEA, PSI 

PSI SCK•CEN SCK•CEN

Program ARIANE ARIANE ARIANE MALIBU MALIBU MALIBU REBUS 

Burnup 

(GWd/MTU) 60.7 52.5 31.1 70.3 50.8 46.0 54.0 

Cs-133 7.8 3.8 3.0 3.9 3.9 4.2 8.0 

Cs-134 -6.7 -10.4 -9.9 -11.5 -14.3 -15.2 

Cs-135 1.6 1.1 9.1 -1.6 -2.8 -0.4 4.4 

Cs-137 1.1 3.6 9.1 -0.4 0.1 -0.3 -1.3 

Ce-144 -0.1 4.4 4.4 -6.3 -8.1 -8.0 -3.7 

Pm-147 39.0 124.8 -61.4 2.0 8.4 -13.2 

Sm-147 -5.7 1.7 6.8 -11.7 -5.9 -2.6 -4.8 

Sm-148 -11.6 -13.1 -3.0 -20.5 -12.4 -8.0 -13.7 

Sm-149 1.6 24.4 7.2 14.8 8.7 11.1 5.1 

Sm-150 5.4 8.3 10.2 -3.9 0.0 4.5 2.2 

Sm-151 34.0 37.2 35.7 31.4 20.4 25.5 36.9 

Sm-152 26.2 38.8 27.9 25.5 24.5 28.6 30.2 

Sm-154 -2.3 6.1 10.8 -4.8 -2.9 -6.8 -1.5 

Eu-151 -42.7 -18.5 43.8 11.8 

Eu-153 11.5 5.4 3.4 9.1 4.1 3.6 6.8 

Eu-154 18.1 -0.6 5.2 11.4 -1.1 -1.8 10.5 

Eu-155 -29.3 -36.5 -32.9 -24.6 -29.4 -38.7 -42.2 

Gd-155 -22.1 -20.9 -51.3 -22.8 -26.2 -38.0 -30.2 

Sr-90 -20.6 3.9 2.1 -1.6 -1.1 -11.7 

Mo-95 -5.4 -2.4 -10.5 0.1 -8.7 -0.8 11.4 

Tc-99 6.4 8.1 30.6 18.6 17.0 -2.6 -2.2 

Ru-101 8.8 0.5 -1.4 12.8 -9.7 11.9 30.5 

Ru-106 9.7 -13.1 -14.9 7.8 -3.2 2.9 

Rh-103 11.3 26.0 -5.7 6.7 -3.8 0.4 23.3 

Pd-105 58.5 

Pd-108 60.1 

Ag-109 118.6 5.7 169.7 137.9 0.3 32.0 

Sb-125 47.1 62.5 63.4 15.7 38.6 

 
43 

Table 5.7. C/E-1 (%) for Vandellós II samples 

Sample ID E58-88 E58-148 E58-260 E58-700 165-2a 160-800 

Burnup

(GWd/MTU) 42.5 54.8 64.6 77.0 78.3 70.9 

U-234 3.9 -6.1 18.9 14.3 16.9 

U-235 -4.2 -14.6 -1.9 -4.8 -1.9 

U-236 8.7 -1.3 3.0 5.6 3.3 

U-238 -0.2 -0.1 -0.4 -0.2 0.4 

Pu-238 0.3 20.4 -3.4 -2.2 -7.6 -9.1 

Pu-239 0.9 4.4 2.3 2.9 -6.0 2.9 

Pu-240 8.3 12.1 6.6 6.6 0.2 0.0 

Pu-241 -0.9 5.6 -0.6 0.9 -7.2 -4.3 

Pu-242 8.8 17.8 6.2 5.7 -1.7 -4.2 

Np-237 -1.8 -11.5 -5.4 0.7 -12.8 -0.2 

Am-241 23.8 29.9 19.0 23.5 10.8 19.7 

Am-243 50.8 25.4 27.4 28.6 -2.8 20.6 

Cm-244 106.9 103.0 79.1 99.6 68.3 62.7 

Cm-246 -10.0 49.7 11.4 34.0 17.3 -14.5 

Nd-142 25.3 34.0 21.9 17.2 16.5 

Nd-143 -0.9 1.8 0.4 1.1 8.0 3.2 

Nd-145 1.2 2.5 0.0 -1.0 4.5 2.6 

Nd-146 2.4 3.9 1.9 1.4 6.0 2.7 

Nd-148 0.4 -0.5 1.3 -2.3 -6.8 1.7 

Nd-150 -0.4 8.0 -0.5 6.7 -6.2 0.9 

Cs-133 3.1 6.1 -3.6 10.6 15.8 

Cs-134 -25.1 -14.3 -15.8 -11.2 -11.8 -12.6 

Cs-135 -10.8 -15.1 -20.5 -2.2 -7.5 

Cs-137 -5.2 0.2 -1.6 0.8 -0.1 -3.1 

Ce-140 1.8 5.3 -0.9 1.5 1.2 -4.4 

Ce-142 -0.4 -0.4 -2.4 -2.1 -2.8 -3.8 

Ce-144 19.0 23.6 21.0 24.7 21.9 

Sm-147 -2.1 -1.5 4.0 -0.6 -10.7 

Sm-148 3.8 2.5 0.7 -4.1 -12.3 -12.2 

Sm-149 -0.1 0.9 12.5 26.6 29.9 16.2 

Sm-150 5.3 3.8 6.1 5.5 9.4 -1.6 

Sm-151 23.8 36.9 38.8 32.7 37.1 30.2 

Sm-152 26.7 30.9 34.2 35.4 30.7 28.5 

Sm-154 6.4 17.9 22.8 13.7 -4.5 -5.2 

Eu-153 -7.1 -3.3 -4.0 2.1 -0.5 

Eu-154 5.6 11.2 18.8 43.1 8.0 

Eu-154 -24.8 -29.2 -26.0 -16.3 -15.6 

Gd-154 41.3 24.4 51.8 37.7 10.5 35.5 

Gd-155 -9.3 -17.1 -7.4 -11.8 6.9 

Gd-156 9.1 5.6 6.3 2.7 2.9 -8.6 

Gd-158 27.4 29.6 39.0 40.7 40.0 25.4 

Gd-160 -6.5 10.8 13.7 15.2 -0.8 -2.6 

Ru-106 -17.3 -6.3 -6.5 -0.5 1.4 

La-139 5.6 11.4 6.3 8.1 

Rh-103 8.0 

Tc-99 6.9 2.9 
44 

5.3 Data Evaluation Criteria 

Prior to applying the experimental data for isotopic validation, the data were examined to identify 

potential systematic measurement problems or isolated errors associated with fuel samples or specific 

isotopes. Systematic biases can occur as a result of radiochemical analysis techniques. The root cause of 

the bias will determine if the bias is correlated with a specific measurement campaign, certain samples, or 

individual isotopes. Biases are also introduced by the computational model and modeling 

approximations, uncertainties in the fuel design description and reactor operating information, and 

deficiencies in nuclear data. Errors associated with the experiment will ultimately be manifest as code 

bias and uncertainty and may lead to inaccurate conclusions about the accuracy of the calculations. 

The aggregate measurement data from each laboratory were examined to assess the general trends and 

behavior of all samples measured at each facility with respect to other laboratories. This assessment 

generally required knowledge of the measurement methods and accuracies. In several cases, laboratory 

results were observed to be clearly different than most other data obtained for comparable fuel samples. 

However, it was not always possible to conclude that these differences were caused by experimental 

problems and not by calculations. When the discrepancies were identified as probable experimental 

errors, further assessment was required to determine if other isotopic data in the same sample or data set 

would be subject to similar issues, or whether the discrepancies were isolated to a single isotope (e.g., 

due to a very low atomic concentration in the fuel). When widespread isotopic inconsistencies were 

observed, the evaluation needs to consider at what point to recommend removal of an entire set of 

measurements. 

The following guidelines were used to evaluate the experimental data and corresponding code-predicted 

data, and also the intrinsic errors and uncertainties in each of these data sets: 

• The results from each laboratory were evaluated for self-consistency in order to identify individual 

discrepant measurement data. 

• The (C/E-1)% comparisons were tested for normality and to verify that the observed variations were 

consistent with the reported experimental errors. 

• Consistency checks were performed for nuclides that are highly correlated at the time of measurement 

by the physics of parent-daughter relationships. For example, the concentration of 241Am is produced 

mainly by the decay of 241Pu, and consequently, the bias for these nuclides are expected to be very 

similar; a similar relationship is expected between 155Gd and 155Eu. 

• Checks were performed to verify that the measured variation in the nuclide concentration as a 

function of burnup was consistent with the physics of isotope generation and depletion. 

• Checks were also performed to examine the variation in the nuclide concentration with the axial 

location for samples obtained from the same fuel rod. 

• Large discrepancies for individual nuclides within a set of laboratory data were examined to 

determine whether the errors were likely caused by isolated measurement problems for specific 

isotopes, or whether systematic deviations were observed for multiple isotopes in the same sample 

that might indicate larger issues with the sample itself. 

• Measurements performed at a specific laboratory were compared with corresponding data from other 

independent laboratories. Agreement, particularly when different analytical methods were used that 

would reduce the likelihood of common bias in the measurements, provided a high level of 

confidence in the results; whereas, results exhibiting a clear and consistent bias with respect to other 

data sets are an indication of likely code model or measurement data issues. 
45 

Isotopic results identified in the review as nonconforming were generally retained in this study. It is 

important not to remove data without clear indication of a measurement error (e.g., on the basis on nonphysical results), as problems may be associated with the computational model or the accuracy of the 

information used in the calculation. Removing potentially reliable measurements on this basis can lead to 

an underestimation of the computational uncertainties. Further evaluations were performed to examine 

and identify the root cause of discrepant data where possible, by looking at all aspects of the fuel 

irradiations, measurements, and calculations. The statistical analysis applied in the evaluations, described 

in the following section, provided an additional measure for assessing data quality. 

5.4 Statistical Analyses 

The results were statistically examined to provide a quantitative assessment of the data. The following 

statistical tests were applied: 

• The mean and standard deviation of the data for each individual laboratory was calculated. 

• The Shapiro-Wilk normality test38 was used to determine whether the isotopic results from each 

laboratory had a normal distribution. 

• The normality test was applied to the combined data from all laboratories and samples. 

• A linear regression analysis was performed to determine the significance of the slope. 

• The sample mean and standard deviation were calculated for the combined data of each isotope. 

• The two-sided tolerance interval within which 95% of the data lie, with a 95% confidence (p = 95%). 

It is important to emphasize that acceptance or removal of data was not based on the results of the 

statistical analyses. It is possible that biases introduced by the models and nuclear data could result in 

statistical anomalies that, if applied in a strict statistical analysis, might lead to removal of good 

measurement data. The measurement data are not obtained from a random sampling of fuels, and many 

of the samples are highly correlated to one another. For example, the 11 samples measured by ANL were 

obtained from the same TMI-1 fuel rod (H6). Any bias attributed to either the calculations or the 

measurements will be highly correlated in these samples and could lead to anomalous distributions of 

results caused by over sampling of data analysed by one laboratory. The statistical tests, however, do 

provide insight into the behavior of the data but alone are not able to account for all of the complex 

relationships between the results obtained from different samples, analytical measurement methods, 

errors, and correlations in the calculations. 

5.4.1 Normality Tests 

The results of the Shapiro-Wilk normality tests are listed in Tables 5.8–5.11 for the isotopes of primary 

interest (see Table 5.1) to spent fuel applications. The tables list the total number of measurements 

available for each isotope and indicate whether the laboratory data and combined data set are normal. A 

large number of actinides and fission products do not meet a p = 95% acceptance criterion for normality 

when all data are included. This condition generally occurred when the results from one or more data sets 

showed large systematic differences with other results. Further evaluations were performed to identify 

problematic data (systematic discrepancies in either the measurements or calculations) that resulted in a 

combined dataset that was not normally distributed. 

Several samples were removed from consideration before performing the normality tests. As discussed in 

more detail in Section 5.5, Takahama-3 samples SF95-1, SF96-1, and SF97-1 were obtained from axial 

rod positions very close to the active end of the fuel, causing results for many nuclides to be 
46 

systematically biased due to spectral effects that could not be accurately characterized by the code 

models. The bias is clearly evident in the isotopic results for plutonium, neptunium, americium, and 

curium, plotted in Figures A.5 – A.18. 

Results obtained from different fuel rods and different laboratories may not necessarily lead to a normally 

distributed data set when combined, even when the results from each laboratory are normal. Laboratories 

and measurement campaigns may be subject to different sources of systematic (nonrandom) bias due to 

variations in fuel sample preparation, separations chemistry, analytical methods, standards, and 

instrumentation errors. Also, the calculations for samples from different fuels may also exhibit 

systematic computational bias due to errors, uncertainties, or approximations in reactor data, operating 

conditions, computational methods, and models. This effect is evident in some of the isotopic results 

plotted in Appendix A. For example, the distribution of data for 240Pu is generally very consistent within 

each reporting laboratory; however, each laboratory set shows a different bias with respect to the other 

measurement sets. Results for 242mAm and 243Am are observed to be normally distributed by laboratory; 

however, the combined data sets are not normal. In general, this can often be attributed to large 

systematic bias in one or more individual data sets. This is not necessarily a definitive indication of 

experimental error, and further evaluation of the data is required. 

The individual laboratory’s results are generally found to have a normal distribution. Some exceptions 

are, for example, the Takahama-3 fuel samples, for which the 237Np and 134Cs results are not normally 

distributed. Further evaluation indicated that the observed discrepancies are due to the data 

corresponding to the SF96 gadolinium poison rod, and the cause is likely associated with the calculations. 

This is discussed further in the next section. 

Laboratory data sets that exhibited a standard deviation significantly larger than the estimated 

measurement error, or showed a clear bias with respect to other data sets in the combined data, were 

examined to identify the possible cause of the large observed variation. 

5.4.2 Linear Regression Analysis 

A linear regression analysis of the calculated and measured data comparisons was performed to identify 

statistically significant trends in the results as a function of sample burnup. The following nuclides have a 

significant (p = 95%) slope: 143Nd, 234U, 236U, 240Pu, 242Pu, 242Cm, 133Cs, 137Cs, 152Sm, 153Eu, 154Eu, and 

155Gd. However, further analysis of the results suggests that some trends are associated with the inclusion 

of measurement data with large biases compared to most other data. For example, removal of the TMI-1 

data measured by ANL produced a non-significant slope in the remaining 240Pu and 242Pu data, i.e., no 

significant trend with burnup. Based on these observations, trending analysis based on statistical 

considerations alone may lead to inaccurate conclusions due to isolated biases in some data. 

5.4.3 Bias and Uncertainty Analysis 

The mean and standard deviation of the results from each analytical laboratory are listed in Tables 5.12–

5.17. The tables show the number of samples measured by each laboratory, average result for each 

laboratory ( x ), and the standard deviation (σ ). Only the isotopes previously identified as being 

important to spent fuel application areas are included in the statistical analysis tables.The sample mean of 

the final combined data set, x , was calculated as 

i

n

i E

C

n

x ∑ ⎟

⎠

⎞ ⎜

⎝

⎛ −= 1 1 , 
47 

where n is the total number of measurements for isotope i, and C/E is the ratio of the calculated-toexperimental isotopic concentration. 

The experimental uncertainties depend on the radiochemical analysis techniques and instrument precision 

and vary considerably between laboratories. Consideration was given to weighting the measurements in 

order to account for the different accuracies of the measurements using a weighting factor 2

i i = /1 σω , 

where σ i is the standard deviation associated with each data set. However, due to the nonstandard 

approach of the different laboratories in reporting uncertainties and the inconsistency between reported 

uncertainties and observed variance of the data, no weighting was performed to avoid potentially 

incorrectly biasing the mean using arbitrary weights. 

The variance of results was calculated as 

∑ ⎥

⎦

⎤ ⎢

⎣

⎡ ⎟ − ⎠

⎞ ⎜

⎝

⎛ − − =

n

i i

x

E

C

n

2

2 1

)1(

1

σ , 

and the relative standard error was estimated as σ . 

The data were also examined to identify potential trends with burnup. The existence of trends in the 

computation bias would mandate a regression analysis rather than determining the mean and variance of 

the results. However, no significant trends with burnup were observed for the results. Trends in the bias 

certainly do exist because the sensitivity of the nuclide concentrations to the neutron cross sections in the 

transmutation chains is not constant as a function of burnup.39 However, in most cases, these trends are 

not statistically significant due to the scatter of the data as a result of experimental and computational 

uncertainties. 

5.4.4 Confidence Intervals 

The expression for the variance assumes that the value of the mean, x , is known precisely. However, in 

practice, the mean is estimated from a finite number of measurements. For small sample sizes, the error 

in the mean can be large, and this error must be included in any realistic estimate of the isotopic 

uncertainty obtained from the experimental data. The tolerance interval is a commonly used statistical 

measure that defines a range that includes a specified fraction of the population with a given likelihood 

(e.g., p = 95% probability). Tolerance limits define the end points of the tolerance interval. Tolerance 

limits can be calculated from tolerance factors, multipliers on the standard deviationσ , to include the 

added uncertainty associated with small sample sizes. 

Two-sided tolerance factors are used to define a range within which a specified fraction of the population 

is estimated to reside; one-sided tolerance factors are used to determine a value above or below which that 

sample population is likely to reside. 
48 

Table 5.8. Normality test results for actinides (U, Pu, Am, Np, Cm) 

Laboratory U-234 U-235 U-236 Pu-238 Pu-239 Pu-240 Pu-241 Am-241 Am-243 Np-237 Cm-24

PNNL + KRI Y Y Y Y Y Y Y Y Y 

JAERI N Y Y Y Y Y Y Y Y N Y 

GE-VNC Y Y Y Y Y Y Y Y Y Y Y 

ANL Y Y Y Y Y Y Y Y Y 

SCK•CENa Y Y Y Y Y Y Y Y Y Y 

Studsvik Y N Y Y N Y N Y Y Y 

Combined set N N N N N N N Y N N Y 

a Includes SCK•CEN, ITU, PSI, and CEA measurements of Gosgen and GKN II fuel samples. 

Table 5.9. Normality test results for fission products (Cs, Ce, Nd) 

Laboratory Cs-133 Cs-134 Cs-137 Ce-144 Nd-143 Nd-145 

PNNL + KRI Y Y Y Y Y 

JAERI N N Y Y Y 

GE-VNC Y Y Y Y 

ANL N Y Y 

SCK•CENa Y Y Y Y Y 

Studsvik Y N Y Y Y Y 

Combined set Y N N Y N N 

a Includes SCK•CEN, ITU, PSI, and CEA measurements of Gosgen and GKN II fuel 

samples. 

Table 5.10. Normality test results for fission products (Sm, Eu, Gd) 

Laboratory Sm-147 Sm-149 Sm-150 Sm-151 Sm-152 Eu-151 Eu-153 Eu-154 Eu-155 Gd-15

PNNL + KRI Y Y Y Y Y Y Y Y 

JAERI Y Y Y Y Y Y 

GE-VNC Y Y Y Y Y Y Y Y 

ANL Y Y Y Y Y Y Y Y 

SCK•CENa Y Y Y Y Y Y Y Y Y Y 

Studsvik Y Y Y Y Y Y Y Y Y Y 

Combined set Y Y N N Y N N N Y Y 

a Includes SCK•CEN, ITU, PSI, and CEA measurements of Gosgen and GKN II fuel samples.
49 

Table 5.11. Normality test results for other fission products (metallics) 

Laboratory Sr-90 Mo-95 Tc-99 Ru-101 Ru-106 Rh-103 Ag-109 Sb-125 

PNNL + KRI Y Y 

JAERI Y Y 

GE-VNC 

ANL Y Y Y Y Y 

SCK•CENa Y Y Y Y Y Y Y 

Studsvik Y Y 

Combined set N Y Y Y N Y Y Y 

a Includes SCK•CEN, ITU, PSI, and CEA measurements of Gosgen and GKN II fuel samples. 

Table 5.12. Calculated-to-experimental differences (%)—Major actinides 

U-234 U-235 U-236 Pu-238 Pu-239 Pu-240 Pu-241 

Reactor Laboratory # x σ # x σ # x σ # x σ # x σ # x σ # x σ # 

Calvert Cliffs PNNL, KRI 0 3 -1.7 0.7 3 2.1 0.3 3 -7.8 1.6 3 4.1 1.9 3 0.3 0.4 3 -1.6 1.3 3 

Takahama-3 JAERI 13 4.0 11.9 13 2.1 1.0 13 -0.6 0.5 13 -3.4 5.5 13 4.5 2.9 13 6.5 1.3 13 -0.5 2.5 13 

TMI-1 GE-VNC 8 0.2 1.3 8 3.5 1.2 8 -3.3 0.5 8 -16.8 3.7 8 2.0 3.9 8 -2.3 1.6 8 -7.6 2.8 8 TMI-1 ANL 11 3.6 3.9 11 4.6 11.0 11 3.3 2.1 11 -2.5 19.4 11 14.9 3.7 11 15.4 6.1 11 4.3 2.8 11 

Gösgen + GKN II SCK•CENa 7 16.7 17.3 7 -0.3 3.9 7 -0.2 1.0 7 -6.1 3.3 7 3.0 4.3 7 3.1 0.7 7 -3.2 2.8 7 

Vandellós II Studsvik 5 9.6 10.5 5 -5.5 5.3 5 3.9 3.7 6 -0.3 10.7 6 1.2 3.7 6 5.6 4.7 6 -1.1 4.4 6 

Combined set 44 5.9 11.1 47 1.5 6.4 47 0.6 2.9 48 -5.7 11.5 48 5.8 6.1 48 6.1 6.9 48 -1.1 4.7 48 

aThe Gosgen and GKN II results measured by SCK•CEN, ITU, PSI, and CEA are grouped in the analysis. 

Table 5.13. Calculated-to-experimental differences (%)—Minor actinides 

Am-241 Am-243 Np-237 Cm-242 Cm-244 

Reactor Laboratory # x σ # x σ # x σ # x σ # x 

Calvert Cliffs PNNL, KRI 3 -7.5 2.7 0 3 9.6 5.1 0 0 

Takahama-3 JAERI 0 13 19.4 6.3 9 25.2 29.4 13 -14.3 17.6 13 2.3 7

TMI-1 GE-VNC 8 3.5 14.4 8 12.5 17.9 8 -3.5 3.4 8 -28.4 9.5 8 -6.8 1TMI-1 ANL 11 -1.3 22.7 11 26.9 24.8 11 4.7 4.7 0 0 

Gösgen + GKN II SCK•CENa 7 5.7 14.1 7 19.9 12.2 6 1.8 18.4 5 -8.6 20.2 7 -5.9 4

Vandellós II Studsvik 6 21.1 6.4 6 25.0 17.2 6 -5.2 5.8 0 0 

Combined set 35 4.5 17.6 45 20.8 16.7 43 6.0 18.3 26 -17.5 17.3 28 -2.4 1

aThe Gosgen and GKN II results measured by SCK•CEN, ITU, PSI, and CEA are grouped in the analysis. 
50 

Table 5.14. Calculated-to-experimental differences (%)—Fission products (Cs, Ce, Nd) 

Cs-133 Cs-134 Cs-137 Ce-144 Nd-143 Nd

Reactor Laboratory # x σ # x σ # x σ # x σ # x σ # 

Calvert Cliffs PNNL, KRI 3 1.9 1.2 3 -13.5 8.1 3 -0.7 0.3 0 3 1.2 0.9 3 -

Takahama-3 JAERI 0 13 -10.1 8.0 13 -3.4 1.9 13 2.6 7.8 13 -1.1 1.2 13 

TMI-1 GE-VNC 0 8 -21.8 1.9 8 -5.0 1.4 0 8 2.5 0.5 8 TMI-1 ANL 0 0 11 -2.1 7.5 0 11 8.2 5.2 11 4

Gösgen + GKN II SCK•CENa 7 4.9 2.1 6 -11.3 3.1 7 1.7 3.6 7 -2.5 5.5 7 1.8 3.4 7 

Vandellós II Studsvik 5 6.4 7.4 6 -15.2 5.1 6 -1.5 2.3 5 22.0 2.2 6 2.2 3.1 6 

Combined set 15 4.8 4.5 36 -14.0 7.2 48 -2.2 4.4 25 5.1 10.9 48 2.6 4.5 48 

aThe Gosgen and GKN II results measured by SCK•CEN, ITU, PSI, and CEA are grouped in the analysis. 

Table 5.15. Calculated-to-experimental differences (%)—Fission products (Pm, Sm)

 Sm-147 Sm-149 Sm-150 Sm-151 Sm-152 

Reactor Laboratory # x σ # x σ # x σ # x σ # x σ

Calvert Cliffs PNNL, KRI 3 -1.5 6.4 0 3 6.5 1.7 3 34.6 7.0 3 26.2 4.0 

Takahama-3 JAERI 5 -1.4 1.4 5 -0.3 6.4 5 5.2 0.7 5 31.9 2.8 5 27.9 3.2 

TMI-1 GE-VNC 8 -3.9 1.3 8 15.9 6.0 8 1.6 0.5 8 26.4 4.5 8 15.4 1.6 TMI-1 ANL 11 11.3 7.4 11 19.1 9.5 11 11.8 6.7 11 43.6 10.6 11 36.4 4.2 

Gösgen + GKN II SCK•CENa 7 -3.2 6.0 7 10.4 7.5 7 3.8 4.8 7 31.6 6.4 7 28.8 4.8 

Vandellós II Studsvik 5 -2.2 5.4 6 14.3 12.6 6 4.7 3.6 6 33.3 5.6 6 31.1 3.3 

Combined set 39 1.2 8.2 37 13.4 10.4 40 6.1 5.6 40 34.4 9.2 40 28.2 8.1 

aThe Gosgen and GKN II results measured by SCK•CEN, ITU, PSI, and CEA are grouped in the analysis.
51 

Table 5.16. Calculated-to-experimental differences (%)—Fission products (Eu, Gd) 

Eu-151 Eu-153 Eu-154 Eu-155 Gd-155 

Reactor Laboratory # x σ # x σ # x σ # x σ # x 

Calvert Cliffs PNNL, KRI 0 3 3.2 0.2 3 2.5 6.8 3 -30.6 1.0 3 -34.6 1

Takahama-3 JAERI 0 0 13 2.5 3.3 0 0 

TMI-1 GE-VNC 8 25.9 8.8 8 -7.4 1.3 0 0 8 -42.8 4TMI-1 ANL 11 -11.3 27.7 11 7.8 3.6 0 11 -56.4 6.1 11 -48.7 7

Gösgen + GKN II SCK•CENa 4 -1.4 37.5 7 6.3 3.1 7 6.0 7.7 7 -33.4 6.1 7 -30.2 1

Vandellós II Studsvik 0 5 -2.6 3.5 5 17.3 15.3 5 -22.4 6.1 5 -7.8 8

Combined set 23 3.4 29.3 34 2.0 6.9 28 6.0 9.3 26 -40.7 15.3 34 -36.2 1

aThe Gosgen and GKN II results measured by SCK•CEN, ITU, PSI, and CEA are grouped in the analysis.

Table 5.17. Calculated-to-experimental differences (%)—Fission products (metallics) 

Sr-90 Mo-95 Tc-99 Ru-101 Ru-106 Rh-103 Ag-109 

Reactor Laboratory # x σ # x σ # x σ # x σ # x σ # x σ # x σ 

Calvert Cliffs PNNL, KRI 3 2.2 1.1 0 3 8.7 4.7 0 0 0 0 

Takahama-3 JAERI 0 0 0 0 13 24.0 26.6 0 0 

TMI-1 GE-VNC 0 0 0 0 0 0 0 TMI-1 ANL 0 11 2.0 4.6 11 -0.3 11.5 11 4.4 6.7 0 11 11.6 6.0 11 111.4 55.1 

Gösgen + GKN II SCK•CENa 6 -4.8 9.4 7 -2.3 7.2 7 10.8 12.0 7 7.6 12.9 6 -1.8 10.5 7 8.3 12.6 6 77.4 73.5 

Vandellós II Studsvik 0 0 2 0 5 -5.8 7.3 1 0 

Combined set 9 -2.5 8.3 18 0.3 6.0 23 4.7 11.3 18 5.7 9.4 24 11.3 24.5 19 10.2 8.7 17 99.4 62.2 

aThe Gosgen and GKN II results measured by SCK•CEN, ITU, PSI, and CEA are grouped in the analysis.
52 

5.5 Review of Laboratory Measurements 

A qualitative review of the results from each laboratory was first performed. This served two purposes: 

• the measurements reported by each laboratory were reviewed for self-consistency; 

• laboratory results were reviewed by group with respect to results from other laboratories that 

measured similar fuel types. 

With the exception of the Gösgen and GKN II samples, most laboratories performed measurements for 

multiple samples obtained from the same fuel rod, having very similar properties. These measurements 

are useful for checking measurement data for self-consistency by analyzing measurements as a function 

of sample burnup and axial position in the fuel rod. For the Gösgen samples, the experimental programs 

used independent laboratory cross checks to confirm the accuracy of the measurements. 

5.5.1 PNNL-KRI 

The Calvert Cliffs ATM-104 samples evaluated in this work were measured at PNNL, and extended 

lanthanide measurements were performed at KRI. All three fuel samples were obtained from the 

MKP109 fuel rod. The PNNL results are generally very consistent and within the range of other 

laboratory results. Lanthanide measurements made at KRI are less consistent and exhibit much larger 

deviations compared to other laboratory data. A notable inconsistency is observed for 149Sm, an 

important burnup credit fission product. The 149Sm concentration does not follow a consistent trend with 

the burnup of the samples and shows high variability when compared to calculations, from about -40% to 

+30%, indicating a serious measurement problem. A review of 149Sm results from other laboratories 

suggests that the KRI result for sample 87-72 (104-MKP109-CC) is likely in error. Similar erratic 

behavior was observed for 151Eu in the 87-72 sample. 

Further review of the experiments indicates that the KRI measurements for nuclides with very low 

isotopic concentrations in the fuel samples are unreliable. Based on this observation, the KRI results for 

149Sm and 151Eu were not considered reliable for use in code validation. 

5.5.2 JAERI 

The JAERI results for the Takahama samples are observed to be very consistent, the exception of the fuel 

rod end samples SF95-1, SF96-1, and SF97-1, excluded from this study. The RSD of the calculated-toexperiment comparisons is 1% for 235U and about 3% for the major plutonium isotopes. The results for 

241Am are very erratic. The JAERI measurement data for all isotopes, except for samarium nuclides, were 

back-calculated to the time of discharge. Since most of the 241Am at the time of measurement is attributed 

to 241Pu decay, the procedure to correct measured 241Am for the large 241Pu contribution introduces 

extremely large errors in the reported 241Am concentration. These additional errors were not completely 

considered and have been dramatically underestimated by the laboratory. Consequently, the 241Am 

measurements are not considered reliable. 

Measurement results for Takahama-3 samples SF95-1, SF96-1, and SF97-1 were excluded from 

consideration in this study because of systematic bias associated with the calculations. These three low 

burnup samples exhibit large systematic discrepancies for many isotopes. It is observed that Pu, Am, and 

Cm isotopes are systematically overpredicted in the current study for these samples, as illustrated in the 

plots in Appendix A. These three samples were located very close to the active fuel ends and were 

exposed to a different neutron spectrum than the rest of the fuel. These samples were previously studied, 
53 

and it was determined that three-dimensional models are required to accurately simulate the spectrum 

near the fuel end region.40 

The SF96 UO2-Gd2O3 rod samples exhibit a larger systematic bias for 237Np and 134Cs than the results for 

SF95 and SF97 rod samples that did not contain gadolinium poison. The bias was not observed for other 

isotopes, suggesting that the source of the bias could be a high sensitivity of the cross sections to changes 

in the neutron energy spectrum caused by the poison or self shielding calculations of resonance cross 

sections in the presence of a strong neutron absorber. The SF96 fuel samples were the only samples 

containing gadolinium poison. The present calculations were performed without spatial subdivision 

(radial rings) of the gadolinium-bearing fuel rod due to limitations of the NITAWL code used to perform 

the resonance self shielding calculations. 

The JAERI measurements did not include separate analysis for any undissolved material. The measured 

metallic fission products 106Ru and 125Sb showed extremely large variability and were overpredicted in the 

calculations, which may indicate incomplete recovery of these isotopes in the fuel dissolver solution. 

5.5.3 ANL 

The TMI-1 samples measured by ANL had relatively large experimental uncertainties compared to most 

other laboratories due to the extensive use of lower-precision ICPMS methods and a systematic bias 

observed between two sets of replicate measurements. The data also exhibited a large variability in 

comparisons of calculations against measurements. However, the observed variability is generally 

consistent with the reported measurement error. 

The results for several nuclides are observed to be systematically biased compared to other laboratory 

data for similar fuels, particularly for the plutonium isotopes. The calculated results for 239Pu, 240Pu, 

241Pu, and 242Pu in the 11 TMI-1 samples measured at ANL exhibit a consistent bias (overprediction) 

relative to the experimental data. In the case of 239Pu, for example, the average bias is about 15% 

(±3.7%); whereas, the results for 239Pu corresponding to all other experimental data sets combined 

(34 samples) have a mean bias of only 2.5% (±4.2%). The RSDs of 240–242Pu data are 6.1%, 2.8%, and 

7.4%, respectively, which is generally consistent with the reported experimental errors. 

The root cause of the positive bias in the plutonium data is unknown. It may be due to experimental 

measurement problems, uncertainties in the sample burnup (discussed below) that are not included in the 

error bars in the plots, and errors in the calculations and nuclear cross-section data, or it may be due to 

fuel configuration or operating conditions that are not accurately documented or modeled in the 

calculations. A potential source of uncertainty, which may have contributed to uncertainties in reactor 

operating conditions, was operational anomalies and fuel failure that occurred during cycle 10.17 Some of 

the affected fuel rods were the source of fuel for the isotopic analysis measurements used in this study, 

and measured rod O12 is known to have breached clad integrity. The Electric Power Research Institute 

(EPRI) performed detailed fuel examinations following the cycle 10 failures to identify the root causes of 

fuel phenomena. The affected fuel was subject to elevated crud buildup on the fuel rods, likely localized 

boiling and precipitation of soluble boron on the fuel rods, and elevated fuel temperatures. Fuel rod 

bowing measurements were also made, and distortions of more than 0.3 mm were measured. However, 

biases were not observed in the TMI-1 measurements made at GE-VNC involving similar fuel that was 

also exposed during cycle 10, suggesting that abnormal operating conditions are not a probable cause of 

the discrepancy, although this cannot be altogether eliminated as a potential source of the errors. The 

large bias in the ANL Pu isotopes is also not observed in any other measurements involving similar fuel, 

indicating that the calculations are an unlikely source of the observed bias. 
54 

In the case of 235U, as illustrated in Figure A.2, a very large variance in the TMI-1 results is observed. 

The difference between calculation and measurement is 4.7 ± 11.0%. The variance is considerably larger 

than that observed in other laboratories and much larger than the reported measurement uncertainty of 

3.7%. Inconsistency of the ANL measurement data is revealed by a plot of the 235U concentration as a 

function of burnup, illustrated in Figure 5.3(a), for the Phase I and Phase II measurements performed at 

ANL on samples from the same H6 fuel rod. These results are based entirely on the experimental data 

and do not involve calculations. The burnup values are those based on the measured 148Nd concentration 

in each sample. The expect result is a gradual decrease in the 235U content with increasing burnup. 

Samples A2, B2, C1, C3, and D2 were measured in Phase I, and samples A1B, B1B, B3J, C2B, D1A2, 

and D1A4 were measured in Phase II. The bias between the Phase I and II results is clearly evident. 

Several samples with nearly identical burnup have measured 235U contents that differ by 10%. However, 

the same measurement data, plotted in Figure 5.3(b) as a function of axial height of the sample in the rod, 

exhibits an expected behavior, with an axial variation in 235U concentration that is consistent with a 

typical burnup profile for a PWR assembly. This result clearly indicates a significant error in the sample 

burnup as determined by 148Nd. 

The reported measurement error of 148Nd is about 7%, confirming that the sample burnup is not known 

precisely. Including the effect of the burnup uncertainty in the plots in Appendix A would significantly 

increase the size of the experimental error bars. The error in sample burnup for the TMI-1 samples 

measured at ANL will increase the calculation errors for all nuclides and result in larger differences 

between calculations and experiment. 

A large number of nuclides from the ANL measurements exhibited significant biases that were identified 

as potential cause of the observed non-normal distribution of the combined data set (10 out of 38 

considered isotopes), an indication of possible systematic problems with the ANL data. The problems 

associated with the lower-precision measurements, uncertainties in the sample burnup, bias observed 

between some Phase I and II results, and the operational uncertainties and fuel failures that occurred 

during cycle 10 could justify not recommending use of the ANL measurements for application to code 

validation. However, in this study, all TMI-1 results were used because of the extensive fission product 

measurements made by ANL that provide an opportunity to evaluate uncertainty for isotopes, which 

previously had very few measurements. 

For burnup credit using both actinides and fission products, inclusion of the ANL data will lead to larger 

uncertainties for many actinides, but this increase in actinide uncertainty will be partially offset by the 

increase in the number of fission product measurements and a reduction in associated uncertainty due to 

the larger sample size. 
55 

40 45 50 55 60

6.0x10-3

6.5x10-3

7.0x10-3

7.5x10-3

8.0x10-3

8.5x10-3

9.0x10-3

9.5x10-3

1.0x10-2

Concentration (g/g 238U)

Burnup (GWd/MTU)

Phase I

Phase II

 (a) 

0 20 40 60 80 100 120 140

6.0x10-3

6.5x10-3

7.0x10-3

7.5x10-3

8.0x10-3

8.5x10-3

9.0x10-3

9.5x10-3

1.0x10-2

Sample location (in)

Phase I

Phase II

Concentration (g/g 238U)

 (b) 

Figure 5.3. 235U concentration for TMI-1 samples measured at ANL (Phase I and II) shown as a 

function of (a) estimated burnup and (b) axial height of the sample in the fuel rod. The stated 

uncertainty in the sample burnup based on measured 148Nd is about 8% (50 ± 4 GWd/MTU), large enough 

to account for the observed deviation from an expected linear decrease in the 235U with increasing burnup. 
56 

5.5.4 GE-VNC 

The TMI-1 fuel samples measured at GE-VNC were made with extensive use of high-precision TIMS 

methods with isotopic dilution. The fuel rods from assembly NJ070G measured by GE-VNC had a lower 

burnup than the ANL samples from assembly NJ05YU and did not include measurements for metallic 

fission products. One of the measured fuel rods, O12, was identified as having failed during cycle 10, and 

the other rods were subject to elevated crud deposits. However, these anomalies do not appear to have 

adversely affected the quality of the data for validation. The uranium and plutonium results are consistent 

with results for other similar fuels, even though samples were obtained from several different fuel rods 

and are, therefore, subject to larger computational uncertainty. The RSD of the 235U results is 1.2% and 

less than 4% for the plutonium isotopes. The plutonium results exhibit a negative bias, although the data 

are generally within the range of other laboratory data. 

The root cause of the differences observed between the TMI-1 results for the samples measured at ANL 

and GE-VNC has not been clearly identified. Assemblies NJ070G and NJ05YU resided in adjacent core 

locations and were therefore likely subjected to similar operating conditions. 

5.5.5 SCK•CEN 

Four samples measured at SCK•CEN included fuel from four different fuel rods, each rod from a different 

assembly irradiated in either the Gösgen or GKN II reactors. Therefore, these samples will not be highly 

correlated to one another compared to other experiments involving more similar fuels. The SCK•CEN 

results are observed to be self-consistent and exhibit variability similar to that seen in other experimental 

results. Sample GU1 was measured at both SCK•CEN and ITU, and cross-check measurements generally 

confirm the accuracy of the reported measurement uncertainties. 

An exception to the consistent behavior is observed for the metallic fission product species that exhibit 

high variability, particularly for 106Ru, 109Ag, and 125Sb. The variance for these isotopes is significantly 

larger than the estimated error of the measurements. The SCK•CEN measurements included separate 

analysis of the undissolved residues. These results point to the significant difficulty in obtaining accurate 

measurements for the metallic species, even when careful analysis of the residues is performed. 

The SCK•CEN fuels measured in the ARIANE program (GU1, GU3 and GU4) were obtained from a 

rebuilt assembly, whereby fuel rods in the assembly were reconfigured at some point during irradiation. 

Although this is identified as a potential source of additional computational uncertainty, the results are 

observed to be well within the range of other commercial UO2 fuels, indicating that any uncertainty 

introduced by the reconfiguration of fuel rods is small. In the case of the GU1 rod, the effect of 

reconfiguration was judged to be sufficiently small that it was ignored. However, the reconfiguration of 

rod GU3/4 was explicitly modeled in this study. Also, the Gösgen samples GGU1, GGU2/1 and GGU2/2 

measured in the MALIBU program were irradiated in a MOX core. Previous studies have indicated that 

this would not have a significant effect on the isotopic predictions, and the results in the current study 

confirm this assessment. 

5.5.6 ITU 

ITU measured Gösgen fuel sample GU4 used in this study. In general, the ITU results are observed to be 

well within the range of other data. ITU used high-precision TIMS measurements with isotopic dilution 

for the uranium and plutonium isotopes, and lower-precision ICPMS methods were used for neodymium 

and most other fission products. A reported measurement uncertainty of 6.7% for the 148Nd concentration 

resulted in large burnup uncertainty and is believed to have contributed to some variability between 
57 

calculations and measurements. An anomalous overprediction of the burnup indicator 137Cs by 9.1% 

suggests that the sample burnup based on 148Nd may be too high (measurement error for 137Cs was 1.5%). 

However, no other systematic discrepancies were observed for any other isotopes that would support a 

burnup discrepancy of this magnitude. Reference measurements for sample GU3 were based on a 

combination of measurement data from ITU and SCK•CEN. ITU measurements of 148Nd in the GU3 

sample were considered out-of-trend in comparisons with independent SCK•CEN measurements. The 

ITU data for neodymium were not recommended by the ARIANE program.20

ITU also performed separate analysis of the metallic elements in undissolved residues. The ITU result for 

99Tc in GU3 is observed to be highly discrepant as compared to the SCK•CEN result. However, the 

results for 106Ru are very consistent, although both results indicate large discrepancies (>80%) compared 

to calculations and results for similar samples made at other laboratories. It is extremely difficult to 

reconcile these observed anomalies. The relatively short half-life of 106Ru (1.023 years) makes the 

calculations highly sensitive to the cooling time of the samples. However, an error in the reported cooling 

time was not corroborated by other nuclide results. 

5.5.7 Studsvik Nuclear AB 

The Vandellós samples measured at Studsvik represent the highest burnup samples evaluated in this 

study. In general, the Studsvik results are observed to be consistent with samples measured at the other 

laboratories. The Studsvik measurements were carried out using ICPMS methods that resulted in larger 

uncertainties in some data. However, in most cases, the reported uncertainty of the measurements was 

consistent with the observed variability in results. Measurement of metallic fission product 103Rh was 

only reported for one sample, 160-800, and 99Tc was reported for two samples, E58-88 and 160-800. 

Measurements were carried out in two experimental phases. The second phase provided measurements 

for one sample and cross-check measurements designed to confirm first phase results and reduce the 

experimental uncertainty. The results for 244Cm were observed to be inconsistent with all other data 

(deviations up to 100%), but the measurements had a large relative standard error of about ±20%. The 

very large uncertainty makes these measurements of limited value for code validation, and the Studsvik 

results for this nuclide are not recommended. 

5.6 Analysis of Selected Isotopic Data 

An evaluation of the isotopic results is discussed for the following element groups: U, Pu, Np, Am, Cm, 

lanthanides (La, Ce, Pr, Nd, Sm, Eu, Gd), Cs, and metallic species (Tc, Mo, Ru, Rh, Ag, Sb, Sr). 

The evaluations are restricted to the isotopes listed in Table 5.1 as being important to burnup credit and 

radiological safety applications. The results discussed in this section refer to the percent differences 

between the calculated and measured isotopic concentrations. 

Comparisons of calculated and experimental data shown in Appendix A include the results for all samples 

included in this study. The plots include the three Takahama-3 samples, identified as SF95-1, SF96-1, 

and SF97-1, that were removed from consideration in the uncertainty analysis because of their proximity 

to the active ends of the fuel rods. The figures include the results for all samples. 

5.6.1 Uranium Isotopes 

Results for the uranium isotopes, 234U, 235U, 236U, and 238U, are plotted in Figures A.1 to A.4. 
58 

5.6.1.1 U-235 

The results for 235U are observed to be predicted on average within about 5% of the measurement for most 

samples. The largest deviations are observed for the TMI-1 samples measured at ANL. The deviations 

are directly attributed to the uncertainty in the experimental determination of the sample burnup caused 

by a large uncertainty in the measured 148Nd content as discussed in Section 5.5.3. The sample burnup 

uncertainty is not included in the error bars shown in the plots. Burnup uncertainty will impact the 

calculated isotopic results by different amounts. Uranium-235 concentration, particularly for high burnup 

fuel, is extremely sensitive to the burnup value. 

Analysis of all measurement data for 235U resulted in a mean bias of 1.6 ± 6.2%. Removal of the ANL 

data from the set would result in a significantly reduced mean and standard deviation: 0.7 ± 3.7%. All 

235U data were included in the uncertainty analysis for this study. 

5.6.1.2 U-238 

Several laboratories (ANL, GE-VNC, and Studsvik) reported measured isotopic data as isotopic mass 

relative to 238U mass in the sample. In these cases, the nuclide concentrations were adjusted to mass 

relative to uranium mass initially present in the fuel (unirradiated) using the combined heavy metal mass 

measured in the sample, adjusted for heavy metal mass loss caused by fission using measured 148Nd as a 

measure of integral fissions. For these measurements absolute 238U mass was not reported. Although it is 

possible to derive 238U content from the other isotopic measurements, such a procedure was not used in 

this study. 

All calculated results for 238U were within 1% of the measurement. The PNNL results for the Calvert 

Cliffs sample showed the largest variations and had the largest uncertainty. The GU1 sample had a 

deviation slightly larger than the reported uncertainty. 

5.6.1.3 U-234 

The discrepancies in the 234U results are significantly larger than for the other uranium isotopes analyzed. 

One possible explanation is uncertainty in the specified initial concentration of 234U in the fuel. The 

precision of the initial uranium isotopic vector is generally not given and may be subject to error since the 

234U concentrations in fresh fuel are very low (typically <0.05 wt % 234U/U). In experiments where the 

initial concentration of the minor isotopes 234U and 236U were not specified, the calculations applied 

approximate values derived from other similar experiments. However, the concentrations of 234U and 

236U in unirradiated fuel are dependent on the origin and enrichment of the fuel, and there is considerable 

uncertainty in derived estimates that may contribute to observed discrepancies. The final 234U content is 

highly dependent on the initial concentration in the unirradiated fuel. 

Therefore, experiments that did not provide measured initial 234U content should not be used to validate 

calculated 234U content in spent fuel. The only experiment that did not document the initial 234U 

concentration was Calvert Cliffs 1. 

5.6.1.4 U-236 

The concentration of initial 236U was only documented for some of the samples considered. It is present 

from the use of some recycled fuel in the fuel manufacturing process, since it does not occur naturally. 

However, the concentration of 236U in spent fuel is not very dependent on the initial concentration. The 

results for 236U are well predicted for most samples. 
59 

5.6.2 Plutonium Isotopes 

The results for the plutonium isotopes are plotted in Figures A.5 to A.9. 

5.6.2.1 Pu-238 

The calculated 238Pu results exhibit an average negative bias. The ANL results have a variance that is 

larger than the reported experimental uncertainty and larger than the variance of other data sets. The 

238Pu measurements are generally made by α-spectrometry (other plutonium isotopes were measured by 

ICPMS) and may be subject to larger experimental errors. As the concentration of 238Pu is very sensitive 

to the fuel burnup, it may have been affected by errors in the sample burnup similar to the 235U case, as 

discussed previously. 

5.6.2.2 Pu-239 

The 239Pu data are generally calculated within the range of experimental error and are overpredicted on 

average. The ANL data are observed to be inconsistent with respect to other data. The variance of the 

ANL results, however, is consistent with the reported measurement uncertainty. The results show a 

positive bias not observed with other data sets. The bias is not explained on the basis of the burnup 

uncertainty because the concentration of 239Pu is not highly sensitive to burnup and all samples are biased 

to a similar extent. All 239Pu measurements were used in the uncertainty analysis in this study because the 

bias could not be clearly associated with experimental errors. However, use of the TMI-1 data measured 

by ANL significantly altered the mean bias and increased the standard deviation of the data as compared 

to the case when the ANL data were not considered. 

5.6.2.3 Pu-240 

The results for 240Pu are very similar to those obtained for 239Pu. The TMI-1 data measured by ANL are 

again observed to be systematically overpredicted as compared to all other laboratory measurements. 

5.6.2.4 Pu-241 

The 241Pu results from each laboratory exhibit very similar behavior to the other plutonium data. 

Calculations are generally within the experimental uncertainties. The ANL results exhibit a positive bias 

with respect to other data, but the scatter is consistent with the reported measurement uncertainty. 

5.6.2.5 Pu-242 

The 242Pu results exhibit larger variability than the other plutonium isotopes. The TMI-1 data measured 

by ANL exhibit large variations, but the results generally lie within the range of other measurements. The 

calculated results for the TMI-1 fuel are consistently underpredicted relative to GE-VNC measurements 

and do not agree within experimental uncertainty with the TMI-1 fuel measured at ANL. 

5.6.3 Neptunium-237 

Neptunium-237 is a minor actinide in burnup credit calculations. Results for 237Np (illustrated in 

Figure A.10) show good agreement between calculations and measurements for all samples, with the 

exception of the Takahama-3 data for the samples from gadolinium-bearing fuel rod SF96. This deviation 

is likely associated with the calculations and the evaluated nucldear data (SF96 is the only measured rod 

that contained gadolinium). The ITU result for sample GU4 and the Studsvik results exhibit a negative 

bias relative to the other laboratory results. 
60 

The discrepancy in 237Np requires further investigation. Since many modern fuel designs make extensive 

use of burnable poison rods, additional work is needed to fully understand the nature of the observed 

errors in predicted 237Np content for gadolinium rods. The systematic deviations cannot be dismissed as 

erratic data. Until the issue is resolved, it is recommended that 237Np data not be used in burnup credit 

calculations for fuel assemblies containing integral burnable poison rods, unless the bias observed in the 

calculated results for the poison rod is conservatively addressed in the uncertainty analysis. 

5.6.4 Americium 

The results for the americium isotopes are plotted in Figures A.11–A.13. 

5.6.4.1 Am-241 

The 241Am results exhibit large deviations that are difficult to evaluate. The concentration of 241Am at the 

time of measurement is attributed almost entirely to formation of 241Am from decay of 241Pu after 

discharge. The Takahama-3 results, as previously indicated, were back calculated to the time of 

discharge, and this procedure introduced large errors in the adjusted measurements for 241Am. The 

Takahama results are therefore not considered for validation in this study. 

The 241Am results for other samples (reported at the time of measurement) are expected to mirror very 

closely the results for 241Pu. However, the 241Am results are observed to exhibit much larger deviations 

than its decay parent. The calculated 241Am concentrations are biased consistently higher than 

measurement for samples measured by SCK•CEN and Studsvik results, by >20%. The bias in the 

Studsvik results may be attributed in part to the relatively large uncertainty in the measurements. The 

SCK•CEN data has much lower reported uncertainty; however, the observed 241Am bias is inconsistent 

with 241Pu, and inconsistent with other laboratory results, CEA and ITU included. 

The discrepancy in the SCK•CEN data for 241Am may be caused by a systematic bias in the measurements 

made using α-spectrometry that results in values of the 241Am content that are about 20% low. 

Additionally, the bias likely extends to other americium isotopes determined by TIMS using the 241Am 

concentration to determine the absolute contents of the other isotopes. It is observed that 242mAm and 

243Am are also overpredicted relative to the SCK•CEN data by about 20%. The GE-VNC and ITU results 

do not exhibit a significant bias, and the agreement between the biases seen for 241Am and 241Pu is a good 

indication of the quality of the GE-VNC and ITU 241Am measurements. 

Better estimates of 241Am uncertainty and bias may be obtained indirectly by using the results for 241Pu, 

which are generally measured with much higher precision. Since 241Am is formed primarily from the 

decay of 241Pu at most cooling times of interest to transportation and interim storage (and at the time of 

measurements), any bias in 241Pu will be reflected directly in the 241Am results. For shorter cooling times, 

however, this cannot be assumed. 

5.6.4.2 Am-243 

The calculated 243Am results for most samples are overpredicted relative to the measurements. The ANL 

results exhibit clear systematic bias between the Phase I and Phase II measurements, as seen in Figure 5.4. 

Calculated results are within the Phase I experimental results, whereas the Phase II results exhibit a 

positive bias of about 50%, indicating significant 243Am measurement problems in Phase II. The TMI-1 

data for 243Am measured by ANL are nevertheless included in this study as the deviations in the ANL 

data are similar to those observed in several other experiments. Similar measurement problems were 

observed for 242mAm. These discrepancies highlight the difficulty in obtaining reliable americium 

measurements. 
61 

0 5 10 15 20 25 30 35 40 45 50 55

-100

-80

-60

-40

-20

0

20

40

60

80

100

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL Phase I Phase II

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

sample #

243Am

(C/E-1) (%)

Figure 5.4. Comparison of measurements and calculations for 243Am. The ANL results for the 

Phase I and II measurements are shown separately. 

5.6.5 Curium Isotopes 

The curium isotopic results are plotted in Figures A.14 to A.19. 

The results for the 242Cm and 244Cm isotopes, both important neutron emitters, exhibit relatively large 

variability, which may be partially attributed to the relatively low accuracy of many curium 

measurements made by α-spectrometry. The observed variance is generally larger than the uncertainties 

cited for the measurements. 

The calculated results for the Studsvik data of 244Cm (shown in Figure A.16) exhibit very large deviations 

compared to the measured values. The large uncertainty of the measurements makes these data of limited 

value for code validation. These results are highly inconsistent with other laboratory measurements of 

similar samples. The Studsvik results for 244Cm were therefore not considered in this study. 

5.6.6 Lanthanides 

Results for the neodymium isotopes 142–148Nd and 150Nd are plotted in Figures A.27–A.33. Other 

lanthanides 144Ce, 147–152Sm, 154Sm, 151Eu, 153–155Eu, 154–156Gd, 158Gd, and 160Gd are plotted in 

Figures A.24–A.26 and A.35–A.50. 

5.6.6.1 Nd-143 

The calculated results for 143Nd, one of the six fission products most important for burnup credit, are 

observed to be predicted within the experimental uncertainty. The calculations are generally within about 

5% of experiment, with the exception of the ANL measurements, which exhibit relatively large scatter 

and are biased high on average. The reported uncertainty of the ANL measurements of 143Nd is ±6%. 
62 

5.6.6.2 Nd-145 

The results for 145Nd are observed to be very similar to 143Nd. The isotope is generally well predicted for 

all experimental data. 

5.6.6.3 Nd-148 

For most samples evaluated in this study, the calculational procedure used a sample burnup value that 

yielded a calculated 148Nd value that was consistent with the measured value. Consequently, the 148Nd 

results for these samples, as shown in Figure A.32, exhibit good agreement between calculation and 

experiment. Larger variations were observed for some samples that used 148Nd and other indicators of 

burnup, such as 137Cs, to estimate the sample burnup. Additional measures of burnup were used in 

instances where there was some indication of 148Nd measurement problems. 

5.6.6.4 Ce-144 

The results for 144Ce are observed to be well predicted in most samples, although the scatter of data is 

relatively large due likely to uncertainties in the measurements. The decay daughter of 144Ce 

(T1/2 = 285 days) is 144Pr (T1/2 = 17 minutes), an important gamma-emitting nuclide and a dominant 

gamma source in shielding for decay times less than about 5 years. 

5.6.6.5 Sm-149 

The samarium isotopes have large absorption cross sections, and 147Sm, 149Sm, 150Sm, 151Sm, and 152Sm 

are important to burnup credit criticality calculations that consider fission product absorption. The results 

for the samarium isotopes are plotted in Figures A.35–A.47. 

The results for 149Sm, one of the six most important fission products in burnup credit calculations that 

credit fission products, are shown in Figure A.36. Calculated concentrations are overpredicted on 

average, and the observed bias appears to be dependent on the measurement laboratory. The 149Sm results 

for the Calvert Cliffs samples are highly discrepant and were not recommended in this study based on the 

observed experimental problems previously discussed. 

5.6.6.6 Sm-147 and Sm-150 

Calculated results for 147Sm and 150Sm are predicted within experimental uncertainty for most samples, 

with the exception of the ANL data, which are biased with respect to the other data sets. However, the 

deviations in the ANL data are generally consistent with the reported uncertainty of the measurements. 

The behavior of the ANL data is similar to that observed for many of the neodymium isotopes; the data 

are clearly biased, resulting in a non-normal distribution of the combined data in some cases, but 

generally, the data exhibit a variance that is consistent with the experimental uncertainty. 

5.6.6.7 Sm-151 and Sm-152 

Results for 151Sm and 152Sm exhibit a positive bias of about 30% for all data sets. The variance in these 

results is also large. 

5.6.6.8 Europium Isotopes 

The europium isotopic results are plotted in Figures A.42–A.45. The 153Eu isotope, and to a lesser extent 

151Eu, are fission products important to burnup credit. Europium-154 (T1/2 = 8.6 years) is moderately 
63 

important to decay heat calculations and is a dominant nuclide in gamma ray shielding in the timeframe 

10 to 30 years after discharge. Results for 155Eu are also indirectly important to burnup credit, because 

155Gd is produced in significant quantities by the decay of 155Eu after discharge. Therefore, any bias in 

155Eu will directly affect 155Gd. 

The 154Eu results are generally calculated within the range of experimental uncertainty, although the 

SCK•CEN, Studsvik, and Calvert Cliffs data exhibit relatively large variance. No large discrepant bias 

was observed in the data sets. 

Calculated 153Eu is generally predicted to within 10% of measurement, although a different bias is evident 

in each of the laboratory data sets. The GE-VNC and Studsvik results are uniformly low, and the KRI, 

SCK•CEN, and ANL results are all high. 

Evaluation of 151Eu, plotted in Figure 5.5, was more problematic because the results for most data sets 

exhibited large deviations. Data from ANL, KRI, and SCK•CEN exhibited inconsistent behavior and 

deviations much larger than the estimated measurement uncertainties. The results from these laboratories 

ranged from about -40% to +30%. The variability is inconsistent with results of all other europium 

isotopes. The concentration of 151Eu in fuel is very low, and this may have contributed to experimental 

errors. The deviations of the ANL and KRI data for samples obtained from the same fuel rod 

measurements are a strong indicator of measurement problems. Based on the erratic behavior of the KRI, 

SCK•CEN, and ANL data, these measurements were not recommended for use in this study. 

5.6.6.9 Gd-155 

The results for the gadolinium isotopes are plotted in Figures A.46–A.50. The calculated results for 155Gd 

are consistently underpredicted relative to measurement for all samples, by about 15 to 50%. The 

concentration of 155Gd at the time of the measurement is attributed entirely to the decay of the parent 

155Eu (T1/2 = 4.75 years). The bias observed for 155Eu is very similar to that of 155Gd, confirming that the 

source of the bias in 155Gd results from errors in the predicted concentration of the 155Eu. 

5.6.7 Cesium Isotopes 

The results for the cesium isotopes are plotted in Figures A.20–A.23. 

5.6.7.1 Cs-133 

The 133Cs isotope is one of the six most important fission products in burnup credit calculations. The 

measured results are observed to be well predicted, within 7% of measurement for all 13 samples. No 

measurements of 133Cs were made by ANL, GE-VNC, or JAERI. 

5.6.7.2 Cs-134 

The calculated results for 134Cs are systematically underpredicted in all samples by about 15–20%, with 

the exception of the Takahama-3 gadolinium poison rod SF96 samples that are not recommended in this 

study. Cesium-134 (T1/2 = 2.06 years) is a dominant nuclide in decay heat and shielding calculations for 

decay times less than about 10 years (see Figure 5.2), where 134Cs can represent more than 25% of the 

decay heat. The production route for 134Cs is from 133Cs neutron capture. 
64 

0 5 10 15 20 25 30 35 40 45 50 55

-80

-60

-40

-20

0

20

40

60

80

Calvert Ciffs-1 - PNNL, KRI

TMI-1 - GE/VNC

TMI-1 - ANL Phase I Phase II

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

(C/E-1) (%)

sample #

151Eu

Figure 5.5. Comparison of measurements and calculations for 151Eu. The ANL results for the 

Phase I and II measurements are indicated separately. 

5.6.7.3 Cs-137 

The isotope 137Cs (T1/2 = 30 years), an important gamma source, is generally predicted within about 5% of 

experiment. The variation of the data is generally consistent with the reported measurement uncertainties, 

with the exception of the ANL measurements that exhibit differences that range from -14% to +17%, 

discrepancies likely associated with uncertainties in sample burnup as previously discussed. 

The 662 keV gamma rays, commonly associated with 137Cs decay, are actually due to the decay progeny 

137mBa (T1/2 = 2.5 minutes). Gamma rays and decay energy associated with 137mBa make this nuclide an 

important decay heat and radiological isotope. The isotopic validation results for 137Cs are directly 

applicable to validating the predictions for 137mBa because these nuclides are in secular equilibrium at the 

time of the measurements. 

5.6.8 Metallic Fission Product Isotopes 

The results for the metallic fission products 90Sr, 95Mo, 99Tc, 101Ru, 106Ru, 103Rh, 125Sb, and 109Ag are 

plotted in Figures A.51–A.58. 

5.6.8.1 Mo-95, Tc-99, Ru-101, Rh-103, and Ag-109 

The nuclides 95Mo, 99Tc, 101Ru, 103Rh, 109Ag are often considered in burnup credit calculations that credit 

fission products. Rhodium-103 is one of the six most important burnup credit fission products. The 

metallic elements are known to be difficult to dissolve in nitric acid (usually employed to dissolve 

uranium fuels) and are commonly associated with undissolved residues in high concentrations. The 

Gösgen and GKN II fuel measurements performed by SCK•CEN, ITU, PSI, and CEA in the 

Belgonucleaire experimental programs evaluated undissolved material separately to ensure recovery of all 

fission product material in the reported concentrations. Other laboratories that reported results for some 

metallic species (PNNL and ANL) did not evaluate undissolved material but also did not indicate the 
65 

presence of any significant quantity of residue in the dissolver. Any significant amount of unaccounted 

material would incorrectly appear as an overprediction by the code. In general, the results for many of 

these metallic species exhibit relatively large deviations, likely to be associated with the complexity and 

difficulty of performing accurate measurements. 

Results for 95Mo are well predicted in all samples. The calculated data are within about 10% of the 

measured values for most samples. 

The variations in the 99Tc data are similar to 95Mo for most samples; however, the ITU result for GU4 is 

clearly discrepant relative to the other data, and one of the TMI-1 samples measured at ANL is also 

discrepant. 

Results for 101Ru exhibit large deviations for the samples measured at SCK•CEN, for which the observed 

variation of the results is about 16%. The REBUS sample is discrepant (30% overpredicted) as compared 

to the other three samples measured at SCK•CEN. The ANL results are well within the range of results 

reported by SCK•CEN and ITU. 

The 103Rh results are overpredicted on average by about 13%. Measurements of 103Rh were only 

performed at ANL, SCK•CEN, and ITU. It was noted that one measurement of 103Rh made by KRI for an 

ATM-106 sample evaluated previously (but not used in this study) indicated an underprediction of 103Rh 

rather than the overprediction seen in the samples evaluated here.41

Results for 109Ag are highly erratic and overpredicted. The results give little confidence in the ability to 

accurately measure 109Ag. The results for samples GU3 and GGU2 are well predicted (<10%), whereas 

GU1, measured at ITU, is overpredicted by more than 100% and the GKN II sample is overpredicted by 

31%. The data for the four samples measured at SCK•CEN have an RSD of 56%. The ANL results are 

overpredicted from 34% to 205%, and the data have a relative standard deviation of 55%. The 

experimental problems with measurements of 109Ag performed at ANL are probably biased because of 

dissolution chemistry problems associated with loss of silver in precipitates, resulting in poor 

reproducibility. 

5.6.8.2 Ru-106 and Sb-125 

Measurements of 106Ru and 125Sb are reported by JAERI and SCK•CEN, and 106Ru was also measured by 

ITU and Studsvik. Although the GU3 and GU4 sample results for 106Ru, measured by SCK•CEN and 

ITU, respectively, are similar, the data are clearly problematic, and there is probably an issue common to 

the samples, error in the measurements or data used in the calculations. These particular measurements 

for samples GU3 and GU4 were removed in this evaluation, although the results for the other samples 

measured by SCK•CEN were retained. 

The Takahama results are also very erratic and exhibit a systematic positive bias consistent with a 

potential loss of the metallic species in the residue. However, the source of the variability is not 

confirmed. The Takahama results for 106Ru and 125Sb were therefore not used because of the erratic 

behavior and variability of the data. 

5.6.8.3 Sr-90 

Strontium-90 (T1/2 = 28.8 years) isotope and its decay progeny 90Y (T1/2 = 2.7 days) are fission products 

important in decay heat calculations, and 90Y is an important radiological gamma source. The results for 

90Sr have important implications on radiological dose assessments involving fission product release. The 

90Sr results are also directly applicable to validating 90Y because these isotopes are in secular equilibrium. 
66 

Only six measurements of 90Sr were available. The result for the Gösgen GU1 sample measured by 

SCK•CEN is low compared to the other two samples, GU3 and GGU2, also measured by SCK•CEN, and 

compared to the results for the three Calvert Cliffs samples measured at PNNL. 

5.6.9 Isotopic Bias and Uncertainty Results 

The mean and relative standard deviation of the validation results for the combined data sets, with the 

discrepant data removed from the population of isotopic data, are listed in Table 5.18. Only the isotopes 

identified as being important to burnup credit, decay heat, and radiation source terms are shown. The 

table lists the total number of measurements used, the sample mean (bias), the RSD between calculation 

and experiment, and the p = 95% confidence interval for each nuclide. 

It is important to emphasize that samples were removed only when there was clear evidence of 

experimental problems that yielded erratic and non-physical measurement results. The following list 

summarizes the experimental data that was not considered in the analysis of isotopic bias and uncertainty. 

• Takahama samples SF95-1, SF96-1, SF97-1 (fuel end region samples) 

• Takahama measurements of 241Am (errors caused by back calculation to discharge) 

• Vandellós results for 244Cm (excessive measurement uncertainty) 

• Calvert Cliffs results for 149Sm (erratic experimental data associated with low isotopic concentrations) 

• Gösgen, GKN II, and Calvert Cliffs results for 151Eu (concentrations near detection limit) 

• TMI-1 measurements of 242mAm (experimental problems) 

• Takahama measurements for 106Ru and 125Sb 
67 

Table 5.18. Summary of analysis results for all samples used in this study 

Nuclide No. of 

measurements

C/E–1

(%) 

σ C/E–1

(%) 

Tolerance interval 

(p = 95%/95%) 

Lower (%) Upper (%) 

U-234 44 5.9 11.2 -21.1 33.0 

U-235 47 1.5 6.4 -13.8 16.8 

U-236 47 0.6 3.0 -6.5 7.8 

U-238 27 -0.2 0.4 -1.2 0.8 

Pu-238 48 -5.7 11.5 -23.1 11.8 

Pu-239 48 5.8 6.1 -8.7 20.4 

Pu-240 48 6.1 6.9 -10.4 22.6 

Pu-241 48 -1.1 4.7 -12.3 10.1 

Pu-242 48 -0.5 7.1 -17.5 16.5 

Np-237 43 6.0 18.3 -38.3 50.3 

Am-241 35 4.5 17.6 39.3 48.3 

Am-243 45 20.8 16.7 -19.4 61.0 

Cm-242 26 -17.5 17.3 -63.2 28.2 

Cm-244 28 -2.4 11.1 -31.0 26.2 

Cs-133 15 4.8 4.5 -8.5 18.1 

Cs-134 36 -14.0 7.2 -31.8 3.9 

Cs-137 48 -2.2 4.4 -12.7 8.3 

Ce-144 25 5.1 10.9 -23.5 33.8 

Nd-143 48 2.6 4.5 -8.1 13.4 

Nd-145 48 1.6 2.4 -4.1 7.3 

Sm-147 39 1.2 8.2 -18.9 21.3 

Sm-149 37 13.4 10.4 -12.3 39.1 

Sm-150 40 6.1 5.6 -7.6 19.8 

Sm-151 40 34.4 9.2 11.9 56.9 

Sm-152 40 28.2 8.1 8.4 48.0 

Eu-151 23 3.4 29.3 -75.0 81.7 

Eu-153 34 2.0 6.9 -15.2 19.3 

Eu-154 28 6.0 9.3 -18.0 30.0 

Eu-155 26 -40.7 15.3 -80.7 -0.7 

Gd-155 34 -36.2 15.9 -74.5 2.1 

Sr-90 9 -2.5 8.38 -32.2 27.2 

Mo-95 18 -0.3 6.0 -17.2 16.6 

Tc-99 23 4.7 11.3 -25.5 34.9 

Ru-101 18 5.7 9.4 -20.8 32.2 

Ru-106 24 11.3 24.5 -53.7 76.3 

Rh-103 19 10.2 8.7 -14.0 34.4 

Ag-109 17 99.4 62.2 -78.4 277.2 

Sb-125 18 75.7 40.0 -37.1 188.5 
68 

5.7 Observations and Recommendations 

Observations from the isotopic data validation performed using the high burnup spent fuel samples 

analyzed in this study are summarized. Several lessons learned from the evaluation process and 

recommendations for future studies that will be performed as new experimental data are made available 

and analyzed are included. 

• There is no physical standard that can be used to calibrate measurement techniques and 

instruments for spent fuel isotopic measurements; the accuracy of measurements can be assessed 

only by careful analysis and intercomparison of data. 

• No single measure or criterion appears to be reliable for assessing experimental data quality. The 

evaluation process requires careful consideration of the analytical techniques used for each 

isotope, and correlations caused by modeling approximations and nuclear data issues, and it 

involves comparison of data with other established measurement laboratories. 

• Statistical tests are useful in evaluating data sets for consistent behavior but alone are not 

sufficient to assess the quality of measurement data. 

• The measurement uncertainties should be used to weight the different measurements to account 

for experimental error in future studies, if possible, provided that realistic uncertainties can be 

assigned to the data. Uncertainties were not used quantitatively in this study because the values 

reported by several laboratories appear to be unrealistically small and did not account for 

uncertainties in all phases of the measurement process. 

• Error in the burnup determined from the measurements should be included in the experimental 

error to provide more realistic error bars on the experimental data. 

• Measurement accuracies for the metallic fission products appear to be problematic. Evaluation of 

undissolved residues for metallic fission products can be important to increase the accuracy for 

these nuclides, as previous studies have indicated that the fraction of these nuclides recovered in 

the dissolver solution can vary widely. However, the results for the metallic species exhibit very 

large variance even for laboratories that separately measured these species in undissolved 

residues; measurement errors for these nuclides are likely to be underestimated. 

• Isotopic data acquired from standard commercial reactor fuel assemblies should be used 

whenever possible. However, the reality is that isotopic measurements on spent fuel have 

frequently not been performed for standard fuel, but for fuels intended for other purposes. In this 

study, measurements for high burnup spent fuel were obtained from 

○ Assemblies that were rebuilt (reconstituted) during irradiation; 

○ Assemblies that experienced abnormal operations and fuel rod failures; and 

○ Assemblies irradiated in reactors operating with MOX fuels. 

An assessment of these fuels indicated that they were highly applicable for validation of standard 

commercial UO2 fuel. The results observed in this study confirm that assessment, although the 

nonstandard fuels likely contributed to additional uncertainty. 
69 

• An important consideration is the number of fuel samples obtained from each rod. While cost 

factors may favor acquiring a large number of samples from a single rod, it must be recognized 

that such samples, alone, do not represent an accurate measure of the uncertainty because the 

computational bias due to reactor operating conditions and position of the fuel rod in the 

assembly are highly correlated. Nevertheless, acquisition and analysis of multiple samples from 

the same fuel rod are desirable for the purposes of experimental error analysis. 

• The validation data analyzed in this study are generally representative of modern fuels. The 

database includes fuel samples covering a broad range of enrichments and burnups that are 

representative of the fuel assembly inventory, a representative selection of rods within an 

assembly (include gadolinium poison rods), different assembly designs, and different reactor 

types. 
70 
71 

6 CONCLUSIONS 

An extensive isotopic validation database has been compiled and analyzed to provide a basis for 

evaluating the uncertainties in calculations involving spent nuclear fuel. The bias and uncertainties in the 

isotopic predictions documented in this report are applicable to establishing margins for uncertainty for 

applications including burnup credit criticality calculations, decay heat calculations, and radiation source 

terms. The calculations described in this report were performed using the two-dimensional 

TRITON/NEWT depletion analysis sequence in SCALE 5.1. This sequence supersedes the onedimensional SAS2H sequence, widely used by industry for spent fuel isotopic calculations. The twodimensional sequence is more rigorous for modeling modern heterogeneous fuel assembly designs and for 

modeling some of the more complex fuel reconfigurations that have occurred in some experimental 

programs. The validation results presented in this study are not directly applicable to validating other 

depletion analysis sequences in SCALE. 

The high burnup spent fuel samples evaluated in this study were selected on the basis of having extensive 

actinide and fission product nuclide measurements. The selected samples, obtained from assemblies 

irradiated in six different PWR plants operated in Germany, Japan, Spain, Switzerland, and the United 

States, have enrichments from 2.6 to 4.7 wt % 235U and cover a large burnup range, from 14 to 

78 GWd/MTU. The database therefore covers fuel with burnup levels up to, and beyond, the current 

licensing peak rod limit of 62 GWd/MTU. 

The isotopic validation data in this report provide the technical basis for performing application-specific 

uncertainty assessments. Realistic estimates of application bias and uncertainty can be obtained by 

identifying the relative contribution of the dominant nuclides to the fuel property being calculated, and by 

propagating the isotopic bias and the uncertainties using either bounding or best-estimate methods. 

The calculated isotopic results are generally found to be within the range of the experimental data for 

many important isotopes in spent fuel. It was observed that the calculation bias does not depend to any 

significant extent on the burnup of the samples, and there are no significant trends observed for very high 

burnup fuel. Although the bias in the calculated nuclide concentrations is not observed to depend to a 

measurable extent on burnup, the bias associated with the application will change as the relative 

importance of different isotopes changes as a function of fuel enrichment, burnup, and cooling time. 

It is recognized that inclusion of some of the more problematic data in this analysis will yield relatively 

large and conservative estimates of isotopic uncertainty. Reducing isotopic uncertainties in the future will 

likely require additional measurements performed using high-precision measurement techniques and 

instruments. As additional isotopic measurement data are acquired, the database evaluated in this study 

will also provide a basis for assessing the quality of new experimental data. 
72 
73 

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A-1 

APPENDIX A

COMPARISON OF ISOTOPIC MEASUREMENTS AND 

CALCULATIONS FOR ALL DATA SETS 
A-2 

0 5 10 15 20 25 30 35 40 45 50 55

-80

-60

-40

-20

0

20

40

60

80

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

(C/E-1) (%)

234U

sample #

0 10 20 30 40 50 60 70 80

-80

-60

-40

-20

0

20

40

60

80

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

(C/E-1) (%)

234U

burnup(GWd/MTU)

Figure A.1. Comparison of measurements and calculated values for 234U.
A-3 

0 5 10 15 20 25 30 35 40 45 50 55

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

sample #

(C/E-1) (%)

235U

0 10 20 30 40 50 60 70 80

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup(GWd/MTU)

235U

(C/E-1) (%)

Figure A.2. Comparison of measurements and calculated values for 235U. 
A-4 

0 5 10 15 20 25 30 35 40 45 50 55

-20

-15

-10

-5

0

5

10

15

20

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

sample #

236U

(C/E-1) (%)

0 10 20 30 40 50 60 70 80

-20

-10

0

10

20

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup(GWd/MTU)

(C/E-1) (%)

236U

Figure A.3. Comparison of measurements and calculated values for 236U. 
A-5 

0 5 10 15 20 25 30 35 40 45 50 55

-4

-3

-2

-1

0

1

2

3

4 Calvert Ciffs-1 - PNNL, KRI

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

238U

(C/E-1)(%)

sample #

0 10 20 30 40 50 60 70 80

-4

-3

-2

-1

0

1

2

3

4 Calvert Ciffs-1 - PNNL, KRI

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

(C/E-1) (%)

burnup(GWd/MTU)

238U

Figure A.4. Comparison of measurements and calculated values for 238U. 
A-6 

0 5 10 15 20 25 30 35 40 45 50 55

-60

-40

-20

0

20

40

60

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

238Pu

sample #

(C/E-1) (%)

0 10 20 30 40 50 60 70 80

-60

-40

-20

0

20

40

60

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup(GWd/MTU)

(C/E-1) (%)

238Pu

Figure A.5. Comparison of measurements and calculated values for 238Pu. 
A-7 

sample #

0 5 10 15 20 25 30 35 40 45 50 55

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

239Pu

(C/E-1) (%)

0 10 20 30 40 50 60 70 80

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup(GWd/MTU)

(C/E-1) (%)

239Pu

Figure A.6. Comparison of measurements and calculated values for 239Pu. 
A-8 

0 5 10 15 20 25 30 35 40 45 50 55

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

sample #

(C/E-1) (%)

240Pu

0 10 20 30 40 50 60 70 80

-40

-35

-30

-25

-20

-15

-10

-5

0

5

10

15

20

25

30

35

40

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup(GWd/MTU)

(C/E-1) (%)

240Pu

Figure A.7. Comparison of measurements and calculated values for 240Pu. 
A-9 

0 5 10 15 20 25 30 35 40 45 50 55

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

241Pu

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup(GWd/MTU)

(C/E-1) (%)

241Pu

Figure A.8. Comparison of measurements and calculated values for 241Pu. 
A-10 

0 5 10 15 20 25 30 35 40 45 50 55

-30

-25

-20

-15

-10

-5

0

5

10

15

20

25

30

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

242Pu

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup(GWd/MTU)

(C/E-1) (%)

242Pu

Figure A.9. Comparison of measurements and calculated values for 242Pu. 
A-11 

0 5 10 15 20 25 30 35 40 45 50 55

-100

-80

-60

-40

-20

0

20

40

60

80

100

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

sample #

(C/E-1) (%)

237Np

0 10 20 30 40 50 60 70 80

-100

-80

-60

-40

-20

0

20

40

60

80

100

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup(GWd/MTU)

(C/E-1) (%)

237Np

Figure A.10. Comparison of measurements and calculated values for 237Np. 
A-12 

0 5 10 15 20 25 30 35 40 45 50 55

-80

-60

-40

-20

0

20

40

60

80

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

241Am

sample #

(C/E-1) (%)

0 10 20 30 40 50 60 70 80

-80

-60

-40

-20

0

20

40

60

80

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup(GWd/MTU)

(C/E-1) (%)

241Am

Figure A.11. Comparison of measurements and calculated values for 241Am. 
A-13 

0 5 10 15 20 25 30 35 40 45 50 55

-160

-140

-120

-100

-80

-60

-40

-20

0

20

40

60

80

100

120

140

160

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

242mAm

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-160

-140

-120

-100

-80

-60

-40

-20

0

20

40

60

80

100

120

140

160

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

burnup(GWd/MTU)

(C/E-1) (%)

242mAm

Figure A.12. Comparison of measurements and calculated values for 242mAm. 
A-14 

0 5 10 15 20 25 30 35 40 45 50 55

-100

-80

-60

-40

-20

0

20

40

60

80

100

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

sample #

243Am

(C/E-1) (%)

0 10 20 30 40 50 60 70 80

-100

-80

-60

-40

-20

0

20

40

60

80

100

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup(GWd/MTU)

(C/E-1) (%)

243Am

Figure A.13. Comparison of measurements and calculated values for 243Am. 
A-15 

0 5 10 15 20 25 30 35 40 45 50 55

-100

-80

-60

-40

-20

0

20

40

60

80

100

Takahama-3 - JAERI

TMI-1 - GE/VNC

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

242Cm

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-100

-80

-60

-40

-20

0

20

40

60

80

100

Takahama-3 - JAERI

TMI-1 - GE/VNC

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

burnup(GWd/MTU)

(C/E-1) (%)

242Cm

Figure A.14. Comparison of measurements and calculated values for 242Cm. 
A-16 

0 5 10 15 20 25 30 35 40 45 50 55

-70

-60

-50

-40

-30

-20

-10

0

10

20

30

40

50

60

70

Takahama-3 - JAERI

TMI-1 - GE/VNC

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

243Cm

sample #

0 10 20 30 40 50 60 70 80

-70

-60

-50

-40

-30

-20

-10

0

10

20

30

40

50

60

70 Takahama-3 - JAERI

TMI-1 - GE/VNC

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

burnup(GWd/MTU)

(C/E-1) (%)

243Cm

Figure A.15. Comparison of measurements and calculated values for 243Cm. 
A-17 

0 5 10 15 20 25 30 35 40 45 50 55

-140

-120

-100

-80

-60

-40

-20

0

20

40

60

80

100

120

140

Takahama-3 - JAERI

TMI-1 - GE/VNC

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

244Cm

sample #

(C/E-1) (%)

0 10 20 30 40 50 60 70 80

-140

-120

-100

-80

-60

-40

-20

0

20

40

60

80

100

120

140

Takahama-3 - JAERI

TMI-1 - GE/VNC

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup(GWd/MTU)

(C/E-1) (%)

244Cm

Figure A.16. Comparison of measurements and calculated values for 244Cm. 
A-18 

0 5 10 15 20 25 30 35 40 45 50 55

-100

-80

-60

-40

-20

0

20

40

60

80

100

Takahama-3 - JAERI

TMI-1 - GE/VNC

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

245Cm

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-100

-80

-60

-40

-20

0

20

40

60

80

100

Takahama-3 - JAERI

TMI-1 - GE/VNC

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

burnup(GWd/MTU)

(C/E-1) (%)

245Cm

Figure A.17. Comparison of measurements and calculated values for 245Cm. 
A-19 

0 5 10 15 20 25 30 35 40 45 50 55

-100

-80

-60

-40

-20

0

20

40

60

80

100 Takahama-3 - JAERI

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

(C/E-1) (%)

sample #

246Cm

10 20 30 40 50 60 70 80

-100

-80

-60

-40

-20

0

20

40

60

80

100 Takahama-3 - JAERI

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup(GWd/MTU)

(C/E-1) (%)

246Cm

Figure A.18. Comparison of measurements and calculated values for 246Cm. 
A-20 

0 5 10 15 20 25 30 35 40 45 50 55

-50

-40

-30

-20

-10

0

10

20

30

40

50 Takahama-3 - JAERI

247Cm

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-50

-40

-30

-20

-10

0

10

20

30

40

50 Takahama-3 - JAERI

burnup(GWd/MTU)

(C/E-1) (%)

247Cm

Figure A.19. Comparison of measurements and calculated values for 247Cm. 
A-21 

0 5 10 15 20 25 30 35 40 45 50 55

-30

-20

-10

0

10

20

30

Calvert Ciffs-1 - PNNL, KRI

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

133Cs

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-30

-20

-10

0

10

20

30

Calvert Ciffs-1 - PNNL, KRI

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

133Cs

Figure A.20. Comparison of measurements and calculated values for 133Cs. 
A-22 

0 5 10 15 20 25 30 35 40 45 50 55

-30

-20

-10

0

10

20

30 Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

(C/E-1) (%)

sample #

134Cs

0 10 20 30 40 50 60 70 80

-50

-40

-30

-20

-10

0

10

20

30

40

50 Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

134Cs

Figure A.21. Comparison of measurements and calculated values for 134Cs. 
A-23 

0 5 10 15 20 25 30 35 40 45 50 55

-30

-20

-10

0

10

20

30

Calvert Ciffs-1 - PNNL, KRI

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

135Cs

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-30

-20

-10

0

10

20

30

Calvert Ciffs-1 - PNNL, KRI

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

135Cs

Figure A.22. Comparison of measurements and calculated values for 135Cs. 
A-24 

0 5 10 15 20 25 30 35 40 45 50 55

-30

-20

-10

0

10

20

30

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

137Cs

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-30

-20

-10

0

10

20

30

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

137Cs

Figure A.23. Comparison of measurements and calculated values for 137Cs. 
A-25 

0 5 10 15 20 25 30 35 40 45 50 55

-10

-5

0

5

10

Vandellos II - Studsvik Nuclear AB

sample #

(C/E-1) (%)

140Ce

0 10 20 30 40 50 60 70 80

-10

-5

0

5

10

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

140Ce

Figure A.24. Comparison of measurements and calculated values for 140Ce. 
A-26 

0 5 10 15 20 25 30 35 40 45 50 55

-10

-5

0

5

10

Vandellos II - Studsvik Nuclear AB

sample #

(C/E-1) (%)

142Ce

0 10 20 30 40 50 60 70 80

-10

-5

0

5

10

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

142Ce

Figure A.25. Comparison of measurements and calculated values for 142Ce. 
A-27 

0 5 10 15 20 25 30 35 40 45 50 55

-50

-40

-30

-20

-10

0

10

20

30

40

50

Takahama-3 - JAERI

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

144Ce

(C/E-1) (%)

sample #

10 20 30 40 50 60 70 80

-60

-50

-40

-30

-20

-10

0

10

20

30

40

50

60

Takahama-3 - JAERI

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

144Ce

Figure A.26. Comparison of measurements and calculated values for 144Ce. 
A-28 

0 5 10 15 20 25 30 35 40 45 50 55

-50

-40

-30

-20

-10

0

10

20

30

40

50

Takahama-3 - JAERI

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

142Nd

sample #

(C/E-1) (%)

0 10 20 30 40 50 60 70 80

-50

-40

-30

-20

-10

0

10

20

30

40

50 Takahama-3 - JAERI

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

142Nd

Figure A.27. Comparison of measurements and calculated values for 142Nd. 
A-29 

0 5 10 15 20 25 30 35 40 45 50 55

-30

-20

-10

0

10

20

30

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

143Nd

sample #

(C/E-1) (%)

0 10 20 30 40 50 60 70 80

-30

-20

-10

0

10

20

30

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

143Nd

Figure A.28. Comparison of measurements and calculated values for 143Nd. 
A-30 

0 5 10 15 20 25 30 35 40 45 50 55

-20

-10

0

10

20

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

sample #

(C/E-1) (%)

144Nd

0 10 20 30 40 50 60 70 80

-20

-15

-10

-5

0

5

10

15

20 Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

burnup (GWd/MTU)

(C/E-1) (%)

144Nd

Figure A.29. Comparison of measurements and calculated values for 144Nd. 
A-31 

0 5 10 15 20 25 30 35 40 45 50 55

-30

-20

-10

0

10

20

30

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

145Nd

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-30

-20

-10

0

10

20

30

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

145Nd

Figure A.30. Comparison of measurements and calculated values for 145Nd. 
A-32 

0 5 10 15 20 25 30 35 40 45 50 55

-20

-10

0

10

20

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

146Nd

sample #

(C/E-1) (%)

0 10 20 30 40 50 60 70 80

-20

-15

-10

-5

0

5

10

15

20

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

146Nd

Figure A.31. Comparison of measurements and calculated values for 146Nd. 
A-33 

0 5 10 15 20 25 30 35 40 45 50 55

-20

-15

-10

-5

0

5

10

15

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

148Nd

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-20

-15

-10

-5

0

5

10

15

20

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

148Nd

Figure A.32. Comparison of measurements and calculated values for 148Nd. 
A-34 

0 5 10 15 20 25 30 35 40 45 50 55

-20

-10

0

10

20

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

(C/E-1) (%)

sample #

150Nd

0 10 20 30 40 50 60 70 80

-20

-15

-10

-5

0

5

10

15

20

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

150Nd

Figure A.33. Comparison of measurements and calculated values for 150Nd. 
A-35 

0 5 10 15 20 25 30 35 40 45 50 55

-50

-40

-30

-20

-10

0

10

20

30

40

50

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

sample #

(C/E-1) (%)

147Pm

0 10 20 30 40 50 60 70 80

-50

-40

-30

-20

-10

0

10

20

30

40

50

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

burnup (GWd/MTU)

(C/E-1) (%)

147Pm

Figure A.34. Comparison of measurements and calculated values for 147Pm. 
A-36 

0 5 10 15 20 25 30 35 40 45 50 55

-50

-40

-30

-20

-10

0

10

20

30

40

50

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

147Sm

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-50

-40

-30

-20

-10

0

10

20

30

40

50

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

147Sm

Figure A.35. Comparison of measurements and calculated values for 147Sm. 
A-37 

0 5 10 15 20 25 30 35 40 45 50 55

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

148Sm

sample #

(C/E-1) (%)

0 10 20 30 40 50 60 70 80

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

148Sm

Figure A.36. Comparison of measurements and calculated values for 148Sm. 
A-38 

0 5 10 15 20 25 30 35 40 45 50 55

-60

-50

-40

-30

-20

-10

0

10

20

30

40

50

60

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

(C/E-1) (%)

sample #

149Sm

0 10 20 30 40 50 60 70 80

-60

-40

-20

0

20

40

60 Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

149Sm

Figure A.37. Comparison of measurements and calculated values for 149Sm. 
A-39 

0 5 10 15 20 25 30 35 40 45 50 55

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

150Sm

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

150Sm

Figure A.38. Comparison of measurements and calculated values for 150Sm. 
A-40 

0 5 10 15 20 25 30 35 40 45 50 55

-100

-80

-60

-40

-20

0

20

40

60

80

100

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

151Sm

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-100

-80

-60

-40

-20

0

20

40

60

80

100

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

151Sm

Figure A.39. Comparison of measurements and calculated values for 151Sm. 
A-41 

0 5 10 15 20 25 30 35 40 45 50 55

-60

-50

-40

-30

-20

-10

0

10

20

30

40

50

60

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

152Sm

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-60

-50

-40

-30

-20

-10

0

10

20

30

40

50

60

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

152Sm

Figure A.40. Comparison of measurements and calculated values for 152Sm. 
A-42 

0 5 10 15 20 25 30 35 40 45 50 55

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

sample #

(C/E-1) (%)

154Sm

0 10 20 30 40 50 60 70 80

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

154Sm

Figure A.41. Comparison of measurements and calculated values for 154Sm. 
A-43 

0 5 10 15 20 25 30 35 40 45 50 55

-80

-60

-40

-20

0

20

40

60

80 Calvert Ciffs-1 - PNNL, KRI

TMI-1 - GE/VNC

TMI-1 - ANL

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

(C/E-1) (%)

sample #

151Eu

0 10 20 30 40 50 60 70 80

-80

-60

-40

-20

0

20

40

60

80

Calvert Ciffs-1 - PNNL, KRI

TMI-1 - GE/VNC

TMI-1 - ANL

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

burnup (GWd/MTU)

(C/E-1) (%)

151Eu

Figure A.42. Comparison of measurements and calculated values for 151Eu. 
A-44 

0 5 10 15 20 25 30 35 40 45 50 55

-30

-20

-10

0

10

20

30

Calvert Ciffs-1 - PNNL, KRI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

153Eu

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-30

-20

-10

0

10

20

30

Calvert Ciffs-1 - PNNL, KRI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

153Eu

Figure A.43. Comparison of measurements and calculated values for 153Eu. 
A-45 

0 5 10 15 20 25 30 35 40 45 50 55

-70

-60

-50

-40

-30

-20

-10

0

10

20

30

40

50

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

154Eu

sample #

(C/E-1) (%)

0 10 20 30 40 50 60 70 80

-80

-60

-40

-20

0

20

40

60

80

Calvert Ciffs-1 - PNNL, KRI

Takahama-3 - JAERI

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

154Eu

Figure A.44. Comparison of measurements and calculated values for 154Eu. 
A-46 

0 5 10 15 20 25 30 35 40 45 50 55

-80

-60

-40

-20

0

20

40

60

80 Calvert Ciffs-1 - PNNL, KRI

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

155Eu

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-80

-60

-40

-20

0

20

40

60

80 Calvert Ciffs-1 - PNNL, KRI

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

155Eu

Figure A.45. Comparison of measurements and calculated values for 155Eu. 
A-47 

0 5 10 15 20 25 30 35 40 45 50 55

-60

-50

-40

-30

-20

-10

0

10

20

30

40

50

60

Calvert Ciffs-1 - PNNL, KRI

Vandellos II - Studsvik Nuclear AB

154Gd

sample #

(C/E-1) (%)

0 10 20 30 40 50 60 70 80

-60

-50

-40

-30

-20

-10

0

10

20

30

40

50

60

Calvert Ciffs-1 - PNNL, KRI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

154Gd

Figure A.46. Comparison of measurements and calculated values for 154Gd. 
A-48 

0 5 10 15 20 25 30 35 40 45 50 55

-80

-60

-40

-20

0

20

40

60

80 Calvert Ciffs-1 - PNNL, KRI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

sample #

(C/E-1) (%)

155Gd

0 10 20 30 40 50 60 70 80

-80

-60

-40

-20

0

20

40

60

80 Calvert Ciffs-1 - PNNL, KRI

TMI-1 - GE/VNC

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

155Gd

Figure A.47. Comparison of measurements and calculated values for 155Gd. 
A-49 

0 5 10 15 20 25 30 35 40 45 50 55

-80

-60

-40

-20

0

20

40

60

80

Calvert Ciffs-1 - PNNL, KRI

Vandellos II - Studsvik Nuclear AB

156Gd

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-80

-60

-40

-20

0

20

40

60

80

Calvert Ciffs-1 - PNNL, KRI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

156Gd

Figure A.48. Comparison of measurements and calculated values for 156Gd. 
A-50 

0 5 10 15 20 25 30 35 40 45 50 55

-120

-100

-80

-60

-40

-20

0

20

40

60

80

100

120

Calvert Ciffs-1 - PNNL, KRI

Vandellos II - Studsvik Nuclear AB

158Gd

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-120

-100

-80

-60

-40

-20

0

20

40

60

80

100

120 Calvert Ciffs-1 - PNNL, KRI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

158Gd

Figure A.49. Comparison of measurements and calculated values for 158Gd. 
A-51 

0 5 10 15 20 25 30 35 40 45 50 55

-60

-50

-40

-30

-20

-10

0

10

20

30

40

50

60 Calvert Ciffs-1 - PNNL, KRI

Vandellos II - Studsvik Nuclear AB

160Gd

sample #

(C/E-1) (%)

0 10 20 30 40 50 60 70 80

-60

-40

-20

0

20

40

60 Calvert Ciffs-1 - PNNL, KRI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

160Gd

Figure A.50. Comparison of measurements and calculated values for 160Gd. 
A-52 

0 5 10 15 20 25 30 35 40 45 50 55

-30

-25

-20

-15

-10

-5

0

5

10

15

20

25

30

Calvert Ciffs-1 - PNNL, KRI

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

90Sr

sample #

(C/E-1) (%)

0 10 20 30 40 50 60 70 80

-30

-20

-10

0

10

20

30 Calvert Ciffs-1 - PNNL, KRI

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

burnup (GWd/MTU)

(C/E-1) (%)

90Sr

Figure A.51. Comparison of measurements and calculated values for 90Sr. 
A-53 

0 5 10 15 20 25 30 35 40 45 50 55

-30

-20

-10

0

10

20

30

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

sample #

(C/E-1) (%)

95Mo

0 10 20 30 40 50 60 70 80

-30

-20

-10

0

10

20

30 TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

burnup (GWd/MTU)

(C/E-1) (%)

95Mo

Figure A.52. Comparison of measurements and calculated values for 95Mo. 
A-54 

0 5 10 15 20 25 30 35 40 45 50 55

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

99Tc

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-40

-30

-20

-10

0

10

20

30

40

Calvert Ciffs-1 - PNNL, KRI

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

99Tc

Figure A.53. Comparison of measurements and calculated values for 99Tc. 
A-55 

0 5 10 15 20 25 30 35 40 45 50 55

-50

-40

-30

-20

-10

0

10

20

30

40

50

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

101Ru

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-50

-40

-30

-20

-10

0

10

20

30

40

50

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

burnup (GWd/MTU)

(C/E-1) (%)

101Ru

Figure A.54. Comparison of measurements and calculated values for 101Ru. 
A-56 

0 5 10 15 20 25 30 35 40 45 50 55

-100

-80

-60

-40

-20

0

20

40

60

80

100

Takahama-3 - JAERI

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

106Ru

sample #

(C/E-1) (%)

0 10 20 30 40 50 60 70 80

-100

-80

-60

-40

-20

0

20

40

60

80

100

Takahama-3 - JAERI

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

106Ru

Figure A.55. Comparison of measurements and calculated values for 106Ru. 
A-57 

0 5 10 15 20 25 30 35 40 45 50 55

-40

-30

-20

-10

0

10

20

30

40

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

103Rh

(C/E-1) (%)

sample #

0 10 20 30 40 50 60 70 80

-40

-30

-20

-10

0

10

20

30

40

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

Vandellos II - Studsvik Nuclear AB

burnup (GWd/MTU)

(C/E-1) (%)

103Rh

Figure A.56. Comparison of measurements and calculated values for 103Rh. 
A-58 

0 5 10 15 20 25 30 35 40 45 50 55

-250

-200

-150

-100

-50

0

50

100

150

200

250

Takahama-3 - JAERI

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

sample #

(C/E-1) (%)

125Sb

0 10 20 30 40 50 60 70 80

-250

-200

-150

-100

-50

0

50

100

150

200

250

Takahama-3 - JAERI

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

burnup (GWd/MTU)

(C/E-1) (%)

125Sb

Figure A.57. Comparison of measurements and calculated values for 125Sb. 
A-59 

0 5 10 15 20 25 30 35 40 45 50 55

-250

-200

-150

-100

-50

0

50

100

150

200

250

300

350

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

109Ag

(C/E-1) (%)

sample #

 

0 10 20 30 40 50 60 70 80

-350

-300

-250

-200

-150

-100

-50

0

50

100

150

200

250

300

350

TMI-1 - ANL

GKN II (REBUS) - SCK/CEN

Gosgen (ARIANE) - SCK/CEN, ITU

Gosgen (MALIBU) - SCK/CEN, CEA, PSI

burnup (GWd/MTU)

(C/E-1) (%)

109Ag

 

Figure A.58. Comparison of measurements and calculated values for 109Ag. 
NUREG/CR-7012 

ORNL/TM-2010/41