NUREG/CR-6972 

ORNL/TM-2008/015 

Validation of SCALE 5 

Decay Heat Predictions for 

LWR Spent Nuclear Fuel 

Office of Nuclear Regulatory Research 
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NUREG/CR-6972 

ORNL/TM-2008/015 

Validation of SCALE 5 

Decay Heat Predictions for 

LWR Spent Nuclear Fuel 

Manuscript Completed: June 2009 

Date Published: February 2010 

Prepared by 

I.C. Gauld, G. Illas, B.D. Murphy, and C.F. Weber 

Oak Ridge National Laboratory 

Managed by UT-Battelle, LLC 

Oak Ridge, TN 37831-6170 

M. Aissa, NRC Project Manager 

NRC Job Code Y6517 

Office of Nuclear Regulatory Research 

iii 

ABSTRACT 

This report documents the validation of ORIGEN-S computer code predictions of decay heat using 

experimental benchmarks involving irradiated nuclear fuels. The experiments include the measurements 

of decay heat for (1) pulse fission irradiations for many fissionable materials in spent fuel for cooling 

times of interest to severe accident analyses (< 105

 s), and for (2) full-length fuel assemblies over longer 

cooling times of importance to spent fuel storage and transportation. The fuel assembly measurements 

evaluated in this study include all previously reported measurements performed in the United States at the 

General Electric Morris Operations facility and at Hanford Engine Maintenance Assembly and 

Disassembly facility in Nevada. Recent fuel assembly decay heat measurements carried out at the 

Swedish Central Interim Storage Facility for Spent Fuel, CLAB, are also included. The purpose of this 

report is to evaluate the computational bias and uncertainty associated with decay heat predictions when 

using the ORIGEN-S code and cross-section libraries generated employing modules and nuclear data 

libraries of the SCALE 5 code system. This validation study includes a broader range of assembly 

designs, fuel enrichments, cooling times, and higher burnup values than has been reported previously. 

The results will be used to develop technical guidance and recommended margins for uncertainty in decay 

heat predicted with ORIGEN-S. 
iv 
v 

TABLE OF CONTENTS 

Page

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

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

LIST OF TABLES....................................................................................................................................... ix

ACKNOWLEDGMENTS ........................................................................................................................... xi

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

1.1 Background ............................................................................................................................... 1

1.2 Objective and Scope.................................................................................................................. 1

2. FISSION EXPERIMENTS ................................................................................................................. 3

2.1 Description of Measurements.................................................................................................... 3

2.1.1 Oak Ridge National Laboratory Measurements ........................................................... 4

2.1.2 Karlsruhe Measurements.............................................................................................. 4

2.1.3 Tokyo University Measurements ................................................................................. 5

2.1.4 Uppsala University Measurements............................................................................... 5

2.1.5 University of Massachusetts, Lowell Measurements ................................................... 5

2.2 Nuclear Decay Data................................................................................................................... 6

2.3 Fission Benchmark Results ....................................................................................................... 9

2.3.1 235U Fission................................................................................................................... 9

2.3.2 239Pu Fission ............................................................................................................... 12

2.3.3 241Pu Fission ............................................................................................................... 14

2.3.4 238U Fast Fission......................................................................................................... 14

2.3.5 233U and 232Th Fast Fission......................................................................................... 15

2.4 Summary of Fission Experiment Results ................................................................................ 17

3. ASSEMBLY MEASUREMENTS.................................................................................................... 19

3.1 Facility Descriptions................................................................................................................ 19

3.1.1 Hanford Engineering Development Laboratory......................................................... 19

3.1.2 GE-Morris Operation ................................................................................................. 21

3.1.3 Swedish Central Interim Storage Facility for Spent Fuel........................................... 24

3.2 Summary of Measurements..................................................................................................... 25

4. FUEL ASSEMBLY DATA .............................................................................................................. 27

4.1 Assembly Designs—U.S. Measurements................................................................................ 27

4.1.1 San Onofre 1 Assemblies ........................................................................................... 27

4.1.2 Turkey Point 3 Assemblies......................................................................................... 33

4.1.3 Point Beach Assemblies ............................................................................................. 37

4.1.4 Cooper, Dresden, and Monticello Assemblies ........................................................... 41
vi 

TABLE OF CONTENTS (continued) 

Page

4.2 Assembly Designs—Sweden Measurements .......................................................................... 49

4.2.1 BWR 8 × 8 Assemblies .............................................................................................. 49

4.2.2 BWR 9 × 9 Assemblies .............................................................................................. 57

4.2.3 BWR SVEA-64 Assemblies....................................................................................... 60

4.2.4 BWR SVEA-100 Assemblies..................................................................................... 63

4.2.5 PWR 15 × 15 Assemblies........................................................................................... 66

4.2.6 PWR 17 × 17 Assemblies........................................................................................... 70

5. COMPUTATIONAL METHODS AND MODELS ......................................................................... 73

5.1 SCALE Depletion Methods and Data ..................................................................................... 73

5.1.1 ORIGEN-S Code........................................................................................................ 73

5.1.2 Depletion Analysis Methodology............................................................................... 74

5.1.3 SCALE Cross-Section Libraries ................................................................................ 77

5.1.4 ORIGEN-S Data Libraries ......................................................................................... 77

5.1.5 ORIGEN-S Cross-Section Library Generation .......................................................... 77

5.2 SAS2H Assembly Models....................................................................................................... 81

5.2.1 Modeling Techniques................................................................................................. 81

5.2.2 BWR Swedish Assemblies......................................................................................... 83

5.3 TRITON Assembly Models .................................................................................................... 85

5.3.1 ORIGEN-S Cross-Section Libraries for U.S. Assemblies ......................................... 85

5.3.2 ORIGEN-S Cross-Section Libraries for Swedish Assemblies................................... 86

6. COMPARISON OF PREDICTIONS AND MEASUREMENTS .................................................... 89

6.1 Assembly Decay Heat Results ................................................................................................ 89

6.2 Error and Uncertainty Analysis............................................................................................... 92 

6.2.1 Statistics for Uncertainty Analysis ............................................................................. 92

6.2.2 Calorimeter Measurement Error................................................................................. 93

6.2.3 Differences Between Calculations and Experimental Data........................................ 97

6.2.4 Sensitivities and Uncertainties in Calculations ........................................................ 105

6.3 SAS2H Analysis Results....................................................................................................... 106

6.4 Effect of Using Assembly Average Burnup and Coolant Density ........................................ 107

7. SUMMARY AND CONCLUSIONS.............................................................................................. 109

8. REFERENCES................................................................................................................................ 111

APPENDIX A SPENT FUEL ASSEMBLY BENCHMARK DATA SUMMARY OF 

EXPERIMENTAL AND CALCULATED RESULTS ..................................................A-1

APPENDIX B ARPLIB INPUT FILE EXAMPLE TRITON INPUT FILES FOR 

SELECTED ASSEMBLIES ........................................................................................... B-1

APPENDIX C SPENT FUEL ASSEMBLY BENCHMARK DATA COMPARISON OF 

TRITON AND SAS2H RESULTS................................................................................. C-1
vii 

LIST OF FIGURES 

Page

Figure 2.1 Measurements of 235U fission decay heat for a fission pulse from 

ORNL, Lowell, and YAYOI experiments compared with calculations ........................... 10

Figure 2.2 Measurements of 235U fission decay heat for an equivalent fission 

pulse from Karlsruhe and Studsvik experiments compared with calculations ................. 11

Figure 2.3 Decay heat for 235U fission from Karlsruhe measurements compared 

with calculations for the experimental 200-s irradiation conditions................................. 11

Figure 2.4 Comparison of 235U fission decay heat for a fission pulse as predicted 

by the ANS-5.1-2005 standard with recommended uncertainties to the 

ORIGEN-S code result ..................................................................................................... 12

Figure 2.5 Measurements of 239Pu fission decay heat for a fission pulse ORNL, 

Lowell, and YAYOI experiments compared with calculations ........................................ 13

Figure 2.6 Comparison of 241Pu thermal fission decay heat for a fission pulse 

measured by ORNL with calculations .............................................................................. 14

Figure 2.7 Measurements of 238U decay heat for a fast fission pulse compared 

with calculations ............................................................................................................... 15

Figure 2.8 Measurements of 233U decay heat (YAYOI fast reactor) for a fast 

fission pulse compared with calculations ......................................................................... 16

Figure 2.9 Measurements of 232Th decay heat (YAYOI fast reactor) for a fast 

fission pulse compared with calculations ......................................................................... 16

Figure 3.1 HEDL Spent Fuel Calorimeter.......................................................................................... 20

Figure 3.2 Schematic of GE-Morris calorimeter system components ............................................... 22

Figure 3.3 Isomeric view of GE-Morris calorimeter.......................................................................... 23

Figure 3.4 Spent fuel assembly calorimeter at the Swedish CLAB facility located 

at poolside and raised to show the assembly loading hatch above the water level ........... 25

Figure 4.1 Layout of the San Onofre Westinghouse 14 × 14 assembly............................................. 31

Figure 4.2 Layout of the Turkey Point Reactor Westinghouse 15 × 15 assembly............................. 34

Figure 4.3 Layout of the Point Beach Reactor Westinghouse 14 × 14 (LOPAR) assembly.............. 39

Figure 4.4 Layout for the GE 7 × 7 fuel assembly ............................................................................. 43

Figure 4.5 GE 7 × 7 fuel assembly dimensions.................................................................................. 44

Figure 4.6 Layout of a BWR 8 × 8-1 assembly.................................................................................. 56

Figure 4.7 Layout of a BWR 9 × 9−5 assembly................................................................................. 59

Figure 4.8 Layout of a BWR SVEA-64 assembly ............................................................................. 62

Figure 4.9 Layout of a BWR SVEA-100 assembly ........................................................................... 65

Figure 4.10 Layout of Ringhals PWR 15 × 15 assembly..................................................................... 69

Figure 4.11 Layout of Ringhals 3 PWR 17 × 17 assembly (with burnable absorbers)........................ 72
viii 

LIST OF FIGURES (continued) 

Page

Figure 5.1 Calculational flow of SCALE depletion analysis (1-D and 2-D) for one time step ......... 76

Figure 5.2 Pin-cell homogenization procedure used in SAS2H 1-D assembly models ..................... 82

Figure 5.3 Cutaway illustration of the ATRIUM-10 assembly layout with large internal 

square water channel......................................................................................................... 83

Figure 5.4 Layout for Type 2 assemblies with three and four BA rods ............................................. 87

Figure 6.1 Measured vs predicted decay heat for all assemblies. The solid line represents 

the line of agreement between predictions and measurements. ........................................ 90

Figure 6.2 Absolute error between calculated and measured decay heat........................................... 91

Figure 6.3 Relative error between calculated and measured decay heat............................................ 91

Figure 6.4 Regression analysis of repeat measurements.................................................................... 96

Figure 6.5 Repeat measurement results at the GE-Morris facility ..................................................... 96

Figure 6.6 Repeat measurements made at the Swedish CLAB facility.............................................. 97

Figure 6.7 Approximate probability density function for error residuals........................................... 99

Figure 6.8 Comparison of standardized sample and normal cumulative distributions ...................... 99

Figure 6.9 Residual error for each measurement facility ................................................................. 100

Figure 6.10 Residual error for CLAB measurements by reactor type................................................ 100

Figure 6.11 Error for GE-Morris measurements by reactor type ....................................................... 102

Figure 6.12 Residual error in specific decay heat for GE-Morris measurements .............................. 102

Figure 6.13 Residual error in specific decay heat for CLAB measurements ..................................... 103

Figure 6.14 Residual error correlation with assembly burnup ........................................................... 103

Figure 6.15 Residual error correlation with assembly cooling time .................................................. 104

Figure 6.16 Residual error correlation with assembly enrichment..................................................... 104

Figure 6.17 Ratio of decay heat calculated using assembly average parameters to that 

obtained using explicit axial calculations for selected Swedish BWR and 

PWR assemblies ............................................................................................................. 108
ix 

LIST OF TABLES 

Page

Table 2.1 Summary of fission experiments used in present decay heat analysis................................ 4

Table 2.2 Summary of fission product decay data sources................................................................. 7

Table 2.3 ORIGEN-S fissionable isotopes with explicit fission product yields ................................. 8

Table 3.1 Summary of evaluated fuel assembly decay heat measurements...................................... 26

Table 4.1 Summary of fuel data for U.S. assembly measurements .................................................. 28

Table 4.2 San Onofre Unit 1 reactor and fuel assembly design information.................................... 30

Table 4.3 Stainless steel composition data for San Onofre fuel assemblies ..................................... 32

Table 4.4 San Onofre Unit 1 reactor and assembly irradiation data ................................................. 32

Table 4.5 Turkey Point 3 reactor and fuel assembly design information ......................................... 35

Table 4.6 Turkey Point reactor and assembly irradiation data.......................................................... 36

Table 4.7 Activated structural components in Turkey Point and Point Beach assemblies ............... 36

Table 4.8 Point Beach reactor and fuel assembly design information.............................................. 38

Table 4.9 Point Beach reactor operating history data ....................................................................... 40

Table 4.10 Cooper, Dresden, and Monticello and fuel assembly design information ........................ 42

Table 4.11 Cooper Station reactor and assembly irradiation data ...................................................... 45

Table 4.12 Dresden 2 reactor and assembly irradiation data .............................................................. 46

Table 4.13 Monticello reactor and assembly irradiation data............................................................. 46

Table 4.14 Activated structural elements in Cooper, Dresden, and Monticello assemblies .............. 47

Table 4.15 Fuel assembly design data for BWR 8 × 8 assemblies ..................................................... 51

Table 4.16 Ringhals-1 reactor and 8 × 8 assembly irradiation history data........................................ 52

Table 4.17 Oskarshamn-2 reactor and 8 × 8 assembly irradiation data.............................................. 52

Table 4.18 Forsmark-1 reactor and 8 × 8 assembly irradiation data................................................... 53

Table 4.19 Forsmark-2 reactor and 8 × 8 assembly irradiation data................................................... 53

Table 4.20 Barsebäck-1 reactor and 8 × 8 assembly irradiation data ................................................. 53

Table 4.21 Barsebäck-2 reactor and 8 × 8 assembly irradiation data ................................................. 54

Table 4.22 Oskarshamn-3 reactor and 8 × 8 assembly irradiation history.......................................... 54

Table 4.23 Other data for BWR 8 × 8 assemblies .............................................................................. 55

Table 4.24 Fuel assembly design data for BWR 9 × 9 assemblies ..................................................... 57

Table 4.25 Forsmark 1 reactor and 9 × 9 assembly irradiation data................................................... 58

Table 4.26 Fuel assembly design data for SVEA-64 assemblies........................................................ 60

Table 4.27 Forsmark-2 reactor and SVEA-64 assembly irradiation data ........................................... 61

Table 4.28 Oskarshamn-2 reactor and SVEA-64 assembly irradiation data....................................... 61
x 

LIST OF TABLES (continued) 

Page

Table 4.29 Fuel assembly design data for SVEA-100 assemblies...................................................... 63

Table 4.30 Oskarshamn-3 reactor and SVEA-100 assembly irradiation data..................................... 64

Table 4.31 Forsmark-3 reactor and SVEA-100 assembly irradiation data ......................................... 64

Table 4.32 Fuel assembly design data for PWR 15 × 15 assemblies.................................................. 66

Table 4.33 Ringhals-2 reactor and PWR 15 × 15 assembly irradiation data ...................................... 67

Table 4.34 Fuel enrichment and initial uranium load for PWR 15 × 15 assemblies .......................... 68

Table 4.35 Fuel assembly design data for PWR 17 × 17 assemblies.................................................. 70

Table 4.36 Ringhals-3 reactor and PWR 17 × 17 assembly irradiation data ...................................... 71

Table 5.1 Cross-section positions and burnup values created in the SCALE libraries..................... 80

Table 6.1 Measurement uncertainty estimates for each calorimeter facility..................................... 95

Table 6.2 Assemblies considered for axial decay heat studies ....................................................... 107

Table A.1 Summary of measured and predicted PWR fuel assembly decay heat 

(U.S. measurements only)...............................................................................................A-3

Table A.2 Summary of measured and predicted BWR fuel assembly decay heat 

(U.S. measurements only)...............................................................................................A-4

Table A.3 Summary of measured and predicted PWR fuel assembly decay heat 

(SKB measurements only) ..............................................................................................A-8

Table A.4 Summary of measured and predicted BWR fuel assembly decay heat 

(SKB measurements only) ............................................................................................A-10

Table C.1 Comparison of TRITON and SAS2H decay heat results for selected 

Swedish PWR fuel assemblies........................................................................................ C-3

Table C.2 Comparison of TRITON and SAS2H decay heat results for selected 

Swedish BWR fuel assemblies ....................................................................................... C-5
xi 

ACKNOWLEDGMENTS 

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

Regulatory Research. Many people made valuable contributions to this work. Measurement data for the 

fission experiments performed at University of Lowell, Massachusetts, were provided by G. Couchel and 

W. Schier; data measured at the University of Tokyo YAYOI fast reactor were provided by J. Katakura 

and Y. Ohkawachi; and J. K. Dickens provided valuable information on the measurements performed at 

Oak Ridge National Laboratory. The results of recent spent fuel assembly calorimeter measurements 

performed at the Central Interim Storage Facility for Spent Fuel in Sweden were made available through a 

technical cooperation agreement between Oak Ridge National Laboratory and the Swedish Nuclear Fuel 

and Waste Management Co., SKB. L. Agrenius, Agrenius Engineering Co., facilitated the exchange of 

data and information required for the computational analysis of the Swedish fuel measurements and 

provided many helpful discussions and guidance. M. McKinnon at Pacific Northwest Laboratory was 

instrumental in making the General Electric Morris operation calorimeter design information available; 

this helped guide the design of the calorimeter constructed by SKB for the Swedish measurements. 

Finally, the authors would like to thank S. M. Bowman, SCALE Project Leader, G. Radulescu and 

B. L. Broadhead for their peer review of this document, C. C. Southmayd for editorial comments, 

A. C. Alford for clerical support, and W. C. Carter for her preparation and formatting of the final report. 
xii 
1 

1. INTRODUCTION 

This report documents the validation of decay heat calculations performed using the ORIGEN-S isotope 

generation and depletion code,1 a module of the SCALE 5 code system,2 against experimental decay heat 

measurements. The measurements include (1) total energy release (decay heat) following fission of 235U, 

239Pu, 241Pu, 238U, 232Th and 233U over very short cooling times of interest to severe accident analysis, and 

(2) calorimetric measurements for more than 160 full-length commercial spent fuel assemblies that cover 

cooling times from several years to about 30 years after discharge from the reactor. 

1.1 Background

An increase in the need for dry storage casks and other independent spent nuclear fuel storage 

installations (ISFSIs) is expected in the next decade as spent fuel storage pools reach their design 

capacities. Onsite dry storage at nuclear power plants is recognized by the utilities as an effective interim 

approach to spent fuel management while a program for long-term disposition of the fuel is implemented. 

One of the new challenges faced by facility designers and regulators is that spent fuel now being 

discharged from nuclear plants has significantly higher burnup and higher initial enrichments than 

previously seen. Advances in fuel design over the past several decades, including use of higher 

enrichment fuels and increased use of burnable poisons, and improvements in core optimization have led 

to discharge assembly burnups of 60 GWd/MTU or more. 

Experimental programs designed to measure spent fuel assembly decay, an important design and safety 

criterion for dry spent fuel storage facilities, were performed in the United States (U.S.) in the 1980s 

using calorimeters operated at General Electric Morris Operation and at the Hanford Engineering 

Development Laboratory. These programs measured approximately 80 assemblies, with assembly 

designs available at the time and fuel with a maximum burnup of 39 GWd/MTU and 27 GWd/MTU for 

pressurized-water reactor (PWR) and boiling-water reactor (BWR) fuels, respectively. The 

measurements have been used to validate computer code predictions3

 that form the technical basis of the 

decay heat values in the U.S. Nuclear Regulatory Commission Regulatory Guide 3.54 on decay heat for 

ISFSIs, revised in 1999.4

 The characteristics of modern fuel however have moved well beyond the 

regime where the computer code predictions of decay heat have been validated and beyond the range of 

the current regulatory guide RG 3.54. 

Recently, an experimental program to measure decay heat for spent fuel assemblies at the Central Interim 

Storage Facility for Spent Fuel (CLAB) located in Sweden was initiated under a project managed by the 

Swedish Nuclear Fuel and Waste Management Company Svensk Kärnbränslehuntering AB (SKB). 

Under this program decay heat measurements have been performed for more than 86 fuel assemblies5

having relatively modern designs, and higher enrichments, burnups, and cooling times than previously 

reported. Oak Ridge National Laboratory (ORNL) has collaborated with SKB to assist in the 

computational analysis of the measurements performed at CLAB using ORIGEN-S and the SCALE code 

system. The new measurements made at the CLAB facility greatly increase the amount of data available 

for code validation and allow the range of code application to be accurately quantified over a wide range 

that includes most modern design fuels. 

1.2 Objective and Scope

The objective of this report is to document the validation of ORIGEN-S over a wide range of spent fuel 

characteristics that includes modern fuel designs, enrichments, discharge burnup, and cooling times. The 

validation includes all earlier calorimeter measurements made at the GE-Morris Operation and Hanford 

facilities, and the newer measurements made at the CLAB facility in Sweden. The measurements provide 
2 

an expanded database of benchmarks necessary to extend the validation range of current regulatory guide 

RG 3.54 to include modern high-burnup fuels. The computational bias and uncertainty associated with 

predictions of assembly decay heat using the ORIGEN-S code of the SCALE 5 code system are 

presented. Fuel design information, plant operating data, and the results of the measurements are 

summarized in this report in sufficient detail to allow independent evaluations of the experiments to be 

performed using other methods and data. Additional and more detailed information is available from the 

primary references cited in this report. 

In addition to the spent fuel assembly benchmarks described, this report includes validation at short decay 

times following fission for many of the major actinides in commercial spent fuel. Although a detailed 

quantitative uncertainty analysis of the fission benchmarks is beyond the scope of this report, the results 

demonstrate the accuracy of the computation methods for a wide range of potential applications at time 

values typically addressed by decay heat standards. 

A summary of the validation studies for the fission experiments (short cooling times) involving individual 

fissionable isotopes is given in Sect. 2. The remainder of the report describes the validation using spent 

fuel assembly measurements. Section 3 provides an overview of the fuel assembly decay heat 

measurements and a brief description of the experimental facilities. Details of the fuel assembly data 

(assembly design, reactor operating conditions, and irradiation history) are presented in Sect. 4. The 

computational methods, nuclear data libraries in the SCALE system, and the models used to generate fuel 

assembly cross sections for the burnup analysis are described in Sect. 5. Finally, a comparison of 

predicted and measured values of decay heat for all evaluated assemblies is presented in Sect. 6 along 

with a statistical analysis of the results and recommended values of bias and uncertainty derived from the 

computational analysis. The appendixes to this report include a summary of all reported decay heat 

measurements and the results of code predictions, and provide examples of the fuel assembly models used 

to generate the cross-section libraries for the ORIGEN-S depletion calculations. 
3 

2. FISSION EXPERIMENTS 

Energy release (decay heat) from the decay of fission products in irradiated fuel is important in the safety 

analysis of nuclear reactors after shutdown. Nearly 7% of the total energy from fission is released by the 

decay of radioactive fission products after the fission process ends. Approximately 25% of the residual 

decay energy from fission is released in the form of beta and gamma rays in the first 10 s after fission, and 

about 50% is released within 100 s of fission. The accumulated fission products continue to release 

energy long after fission ends; this energy source plays an important role in the evaluation of postulated 

loss-of-coolant accidents and emergency core cooling system performance that considers a time range of 

less than about 105

 s after fission. 

The rate of energy dissipation over short times after fission has been measured extensively. Because the 

measurement of full-length fuel assemblies at such short cooling times is impractical, these experiments 

generally involve a small sample of fissionable isotope irradiated in a neutron flux for relatively short 

periods of time. These experiments are often referred to as “pulse” fission experiments and have been 

widely used in fission product decay heat standards development. 

These decay heat measurement methods generally fall into two categories: spectroscopic and calorimeter. 

The spectroscopic method involves measuring the gamma and beta radiation spectra using detectors, 

unfolding the spectra to obtain the emitted energy distribution, and deriving the decay heat from the 

individual gamma and beta energy components. Evaluations of the measurements6,7 have found a 

consistent and systematic bias between the spectroscopic and calorimeter results at some decay times that 

has not been resolved. The calorimeter measurements are observed to yield systematically larger results 

than the spectroscopic measurements by upward of 2%, or about one standard deviation assignment to the 

calorimetric results, in the timeframe of t < 100 s after fission; the differences though are in general 

statistically insignificant at the 95% confidence level.6

 Any evaluation of the decay heat after fission and 

the assigned uncertainty based on experimental data will therefore be influenced by the selection of the 

experiments used in the analysis. 

2.1 Description of Measurements

The benchmark measurements selected for this work are listed in Table 2.1. They include measurements 

performed after 1980 of energy release from fission of 233U, 235U, 238U, 239Pu, 241Pu, and 232Th. Only those 

experiments reporting measurements of total decay heat or reporting both the beta and gamma 

components were considered, because calculations were performed only for the total decay heat. These 

experiments used mostly spectroscopic measurement methods and involved relatively short irradiation 

times with respect to the decay times at which the decay heat was measured. The latter feature makes 

these experiments highly relevant to the direct evaluation of decay heat from a pulse fission event at short 

cooling times. The results using spectroscopic techniques yield very consistent agreement with one 

another. However, it must be emphasized that for some time regimes the spectroscopic results are lower 

than corresponding calorimeter results and tend to be in better overall agreement with the calculations 

presented here. A comprehensive review of other experiments performed predominantly before 1980 can 

be found in publications by Tobias,8

 Schrock,9,10 and Dickens.7 The experimental results evaluated in this 

study also tend to be lower than the values obtained from the American National Standard ANSI/ANS-5.1 

on decay heat.11 The values in the Standard are intentionally developed to be conservative and are not 

based on any single set of measurements or even a least-squares best estimate evaluation of more than one 

experiment. Consequently, the ANS-5.1 values are more likely to overpredict than underpredict values 

for the decay heat. The most recent measurements, performed circa 1997, were made at the University of 

Massachusetts at Lowell. The Lowell data have not been included in the development of the 
4 

ANS-5.1 Standard. These measurements are notable because they include decay times < 1 s after fission, 

well below most other measurements. 

Table 2.1 Summary of fission experiments used in present decay heat analysis 

Data 

set Isotopes Methoda

 Author(s) Institute Pub. year

(circa) 

1 235U, 239Pu, 241Pu γ, β spec. Dickens et al. Oak Ridge National 

Laboratory 

 1980, 

1981 

2 235U calorimeter Baumung Karlsruhe 1981 

3 233U, 235U, 238U, 

239Pu, 232Th 

γ, β spec. Akiyama et al. Tokyo University 1982 

4 235U γ, β spec. Johansson Uppsala University/ 

Studsvik 

1987 

5 235U, 238U, 239Pu γ, β spec. Schier and 

Couchell et al. 

University of 

Massachusetts, 

Lowell 

1997 

a

 spec = spectroscopic method. 

2.1.1 Oak Ridge National Laboratory Measurements 

Oak Ridge National Laboratory (ORNL) performed measurements of gamma-ray and beta decay energy 

release following thermal fission of 235U, 239Pu, and 241Pu at the Oak Ridge Research Reactor.12,13 

The gamma measurements were made with a NaI(Tl) scintillation detector. Beta measurements were 

made using a NE-110 plastic beta ray scintillator with gamma coincidence. The results were 

computationally unfolded using the detector response functions to obtain spectral distributions of 

moderate resolution. The resulting energy spectra were then integrated to obtain the total energy released. 

Decay heat values (gamma and beta components) were measured at time intervals from 2 to 14,000 s 

after fission. The decay heat was determined from the integrated detector counts over each counting time 

interval. The irradiation time of the samples was varied to obtain an adequate number of integral counts 

at each of the decay times considered. 

The absolute number of fission events was determined experimentally by separately counting the samples 

after irradiation in a low-background counting room using a high-resolution Ge(Li) detector. 

Characteristic gamma emissions for the fission products 99Mo, 132Te, 143Ce, and 97Nb-97Zr were selected to 

determine the integral fissions that occurred in each sample. All results were reported on a per fission 

event basis. The decay heat values were corrected for loss of noble gases from the samples after fission 

by experimentally measuring the rate of 88Kr diffusion from irradiated samples. 

2.1.2 Karlsruhe Measurements 

Calorimetric measurements for 235U fission by thermal neutrons were performed at Karlsruhe and 

reported by Baumung in 1981.14 Small sample pellets were irradiated for 200 s in the thermal column of 

the Karlsruhe research reactor. The decay heat was measured by the temperature rise on an adiabatic 

calorimeter and corrected for the fraction of gamma ray energy that escaped the calorimeter as determined 
5 

by detector measurement. The fraction of gamma energy escaping the pellet was reported as about 30 to 

40%, depending on the sample size. The escaping gamma energy was measured seperately and added to 

the energy measured by the calorimeter to obtain the total decay heat. The decay heat was reported for 

decay times from 15 to 4000 s after fission. 

Many of the reported calorimetric measurements of decay heat involve sample irradiation times that are 

large relative to the decay times. Consequently, the measurements are not representative of a pulse 

fission event. The calculations for the Karlsruhe experiment were performed for the experimental 

conditions of 200-s irradiation. The measured decay heat values were subsequently adjusted to an 

equivalent fission pulse using conversion factors calculated from the ratio of the decay heat for pulse 

fission and experimental conditions, for the purpose of comparing the Karlsruhe results with the other 

experimental data. 

2.1.3 Tokyo University Measurements 

Experiments to measure gamma and beta decay following fission were performed at the University of 

Tokyo fast research reactor, YAYOI.15–17 These experiments, referred to here as the YAYOI 

experiments, were very similar to the ORNL measurements in terms of the methodology. However, they 

included measurement of several additional fissionable materials for fast neutron induced fission. 

The measurements reported here include fission of 233U, 235U, 238U, 239Pu, and 232Th. The difference 

between the decay heat from fast neutron induced fission of 235U and 239Pu is estimated to be less than 1% 

as compared with that of thermal neutron induced fission.7

 This difference is considerably smaller than 

the experimental uncertainties, and the fast fission data are therefore generally considered applicable for 

thermal fission benchmarking purposes. 

2.1.4 Uppsala University Measurements 

Additional independent measurements of 235U decay heat following fission carried out using the 

spectroscopy method have been reported by the University of Uppsala at the Studsvik research center in 

Sweden.18 The measurements were similar to the spectroscopic methods described previously. The beta 

ray measurements were made with a Si(Li) detector system, and the spectra were computationally 

unfolded using the measured energy response of the detector. The gamma ray detection system consisted 

of a 12.5 cm × 12.5 cm NaI(Tl) scintillator. Absolute number of fissions in each sample were determined 

experimentally using fission chambers and a Ge(Li) gamma detector system. 

The measurements were performed for 235U and 239Pu thermal fission and 238U fast fission (2.3 MeV). 

Fast neutrons were produced by a Van de Graff accelerator, and in the case of thermal irradiations, 

thermal neutrons were produced with a paraffin moderator. After the samples were irradiated they were 

transferred to the beta and gamma detectors where the spectra were measured over 16 consecutive time 

intervals from 10 s to 10,000 s after fission. At this time, results for both gamma and beta decay heat 

measurements have only been reported for 235U fission. 

2.1.5 University of Massachusetts, Lowell Measurements 

The University of Massachusetts at Lowell has measured the gamma and beta energy release for the 

fission of 235U, 239Pu, and 238U at decay times from about 0.2 s to 40,000 s after fission (referred to here as 

the Lowell measurements).19 The thermal neutron irradiations were conducted at the University of 

Massachusetts Lowell 5.5-MV Van de Graff accelerator using thermal neutrons from the 7

Li(p,n)7Be 

reaction to induce fission in 235U and 239Pu. Fast neutrons from the 1-MW pool-type research reactor 

were used to induce fissions in 238U. The gamma spectrometer consisted of a 5 × 5 in. NaI(Tl) 

scintillation detector in coincidence with a thin beta detector. The beta spectrometer was a 3 × 3 in. 
6 

plastic scintillation detector gated by a thin surface scintillator. The detectors covered an energy range of 

0.1 to 8 MeV. The measured gamma and beta spectra were unfolded using the detector energy response 

functions to obtain the actual energy spectra from which average energies and total energy release were 

deduced. 

One of the key features of the Lowell measurements is the very short decay times; in many cases less than 

1 s after fission. The number of gamma rays and the gamma energy per beta particle were used to 

normalize the measurements using calculated values of gamma decay heat. This normalization was 

necessary because the absolute number of fissions in the samples was not measured. The measurements 

were also subject to loss of noble gases from the irradiated samples during counting. The loss of noble 

gases and their decay progeny from the samples was corrected using factors calculated from the total 

decay heat both with and without the noble gas contribution. At the longer cooling times of the 

experiment, this correction can become substantial, and the results are therefore influenced by the 

accuracy of the calculation. Consequently, the measurements are not considered as reliable at longer 

decay times as at the shorter decay times where the correction factor is smaller. Although the 

normalization of the measurements and the noble gas correction using code calculations limits the value 

of the Lowell measurements as an absolute benchmark to some extent, the very short decay times of the 

measurements (where the largest fraction of energy is released) make these measurements attractive for 

code validation. 

Gamma and beta decay heat results were provided in tabular format by G. Couchell and W. Shier at 

Lowell.20 The counting intervals for the gamma measurements were different than those for the beta 

measurements. The results of the beta measurements were therefore interpolated to the times of the 

gamma measurements, and the total decay heat was then obtained by combining the gamma and beta 

results. The additional uncertainty introduced by interpolation was estimated to be less than 2% in the 

total decay heat values. 

2.2 Nuclear Decay Data

Decay heat calculations were performed for each of the fission benchmark problems described in 

Table 2.1 using the ORIGEN-S code and associated nuclear data libraries. The accuracy of the decay 

heat predictions for the fission experiments is dependent primarily on the fission yields, the decay 

schemes (metastable levels, decay modes, branching fractions, and half lives) and the recoverable energy 

per decay (Q values) for each fission product. The fission yields implemented in the ORIGEN-S data 

library are derived from ENDF/B-VI. Independent yields for the nuclides 74mAs, 85mSe, 86mBr and 162mTb 

were combined with the yields to their ground states. A review of the 1999 Table of Isotopes21 indicated 

that these metastable states have been eliminated in the current decay scheme evaluations. Furthermore, 

yields to the second metastable states of 116nIn, 118nIn, 120nIn, 122nIn, 124nSb, 126nSb, 130nIn, 152nPm and 152nEu 

were combined with their decay progeny because ORIGEN-S currently does not support nuclide 

identifiers for excited levels beyond the first metastable state. The decay energy was assigned to the 

daughter in such a way as to mitigate the effects of this approximate representation on the decay heat 

predictions. 

The nuclear decay data and decay schemes are also primarily from ENDF/B-VI. Decay data were 

compiled for 1093 individual fission products having direct fission yields from ENDF/B-VI. 

Approximately 230 fission products with direct yields were not included for lack of evaluated decay 

schemes. Most of these were low yield isotopes located far from stability. The sources of nuclear decay 

data in the ORIGEN-S library are listed in Table 2.2. 
7 

Table 2.2 Summary of fission product decay data sources 

Number of 

nuclides Stability, status, and data source 

146 Stable nuclides 

764 Radioactive, ENDF/B-VI decay data 

172 Radioactive, ENSDF decay data 

11 Radioactive, JEF-2.2 decay data 

26 Ternary fission products and decay progeny (JEF-2.2) 

 

1119 Total fission product library size 

Fission product yields for thermal-neutron induced fission were implemented for all materials evaluated 

in this study except 238U and 232Th, for which fast fission yields were applied (represented in ENDF/B-VI 

as neutrons having an incident energy En=500 keV). The actinides in the ORIGEN-S library with explicit 

fission product yields and the corresponding energy of neutron-induced fission are listed in Table 2.3. 
8 

Table 2.3 ORIGEN-S fissionable isotopes with explicit fission product yields 

Number Fissionable 

isotope 

ENDF/B-VI 

MAT number 

ORIGEN-S 

ID 

Fission 

energya

1 227Th 9025 902270 thermal 

2 229Th 9031 902290 thermal 

3 232Th 9040 902320 fast 

4 231Pa 9131 912310 thermal 

5 232U 9219 922320 thermal 

6 233U 9222 922330 thermal 

7 234U 9225 922340 fast 

8 235U 9228 922350 thermal 

9 236U 9231 922360 fast 

10 237U 9234 922370 fast 

11 238U 9237 922380 fast 

12 237Np 9346 932370 thermal 

13 238Np 9349 932380 fast 

14 238Pu 9434 942380 fast 

15 239Pu 9437 942390 thermal 

16 240Pu 9440 942400 thermal 

17 241Pu 9443 942410 thermal 

18 242Pu 9446 942420 thermal 

19 241Am 9543 952410 thermal 

20 242mAm 9547 952421 thermal 

21 243Am 9549 952430 fast 

22 242Cm 9631 962420 fast 

23 243Cm 9634 962430 thermal 

24 244Cm 9637 962440 fast 

25 245Cm 9640 962450 thermal 

26 246Cm 9643 962460 fast 

27 248Cm 9649 962480 fast 

28 249Cf 9852 982490 thermal 

29 251Cf 9858 982510 thermal 

30 254Es 9914 992540 thermal 

a Neutron energy causing fission: thermal = 0.0253 eV; fast = 500 keV. 
9 

2.3 Fission Benchmark Results

The measured and calculated decay heat results are presented as a function of time after fission, in units 

of MeV/fission. Decay energy decreases after fission approximately as 1/t, where t is time after a fission 

event. A common convention used to present these data is to multiply the energy release rate per fission 

(MeV/s per fission) by the decay time (s) to obtain the units of MeV/fission reported in the studies. The 

calculated energy release rates after fission for 233U, 235U, 238U, 239Pu, 241Pu, and 232Th are compared with 

measurements in the following sections. 

2.3.1 235U Fission 

Decay heat from thermal fission of 235U was measured in all of the selected experiments listed in 

Table 2.1. All but one of the measurements (Karlsruhe) were made by gamma and beta ray spectroscopy 

of the particles emitted from the irradiated samples. The calculated results are compared separately to the 

ORNL, Lowell, and YAYOI measurements in Figure 2.1, and to the Karlsruhe and Studsvik 

measurements in Figure 2.2. The measurements are presented separately because the time range of the 

data in Figure 2.2 is much less than the data in Figure 2.1. 

The Karlsruhe and Studsvik results are observed to have somewhat larger uncertainties in the decay time 

range of 10 s to 100 s after fission compared with the other results and in general are seen to yield slightly 

larger estimates of decay heat than the other measurements. The Lowell results for the longer cooling 

times are observed to be significantly lower than all other measurements; this is an effect due likely to 

errors introduced by the calculations to correct the measurements for noble gas loss; this effect becomes 

more important at longer decay times. 

The Karlsruhe results shown in Figure 2.2 are presented for a basis of one fission pulse. However, the 

experiments actually involved an irradiation time of 200 s. Therefore, the results, particularly at short 

decay times, are clearly not representative of a fission pulse. For the purposes of comparison with other 

experimental data, the Karlsruhe results in Figure 2.2 were obtained by adjusting the measured data to an 

equivalent fission pulse using the conversion factor 

η(t) = f(t) / F(t), (2.1) 

where 

 f(t) = calculated decay heat for a fission pulse event, 

 F(t) = calculated decay heat per fission for the experimental irradiation times. 

No additional uncertainty was assigned to the Karlsruhe data to cover the calculational uncertainty. 

The conversion factors (η) ranged from about η = 0.2 at 15 s after fission, to about η = 0.97 at 4000 s 

(indicating that at 4000 s the experiment very nearly simulates a pulse fission). The Karlsruhe data are 

separately compared with the calculations for the actual experimental conditions in Figure 2.3, i.e., 

without correction factors. The results are consistent with those of Figure 2.2 in that the calculations 

exhibit a small negative bias, but are within the stated uncertainty of the measurements over all decay 

times. The agreement suggests that the error associated with η is relatively small. 

Overall, the agreement between calculation and experiment is seen to be within the level of the 

measurement uncertainties. It is important to emphasize here again that many of the calorimeter 

measurements not included in this study (pre-1980) yield decay heat estimates that are larger than the data 
10 

sets selected in this work and that inclusion of these earlier data would likely indicate a larger deviation 

between prediction and measurements than is indicated by the current results. 

A comparison of the decay heat calculated using ORIGEN-S and the ANS-5.1-2005 Standard values for a 

235U fission pulse11 is illustrated in Figure 2.4. The decay heat predicted by ORIGEN-S is observed to be 

less than the Standard and outside the one standard deviation of the uncertainty assigned to the Standard 

over the time range < 103

 s. The differences are seen to be largest in the time range of about 5 to 100 s. 

The differences largely reflect the different experiments used in the developlement of the standard. 

Because the standard conservatively reflects differences between experimental data sets used in the 

development, it will understandably yield decay heat predictions that are larger than those obtained using 

the best estimate values that have not applied any margin for uncertainty. The Standard is developed to 

be conservative and is thus weighted toward those experiments with the larger values of decay heat. For 

longer cooling times corresponding to spent fuel storage, actinides can contribute a significant fraction of 

the total decay heat. Because the Standard currently does not include a comprehensive actinide treatment, 

it can not be used for spent fuel storage applications without supplementary actinide methods or data 

provided by the user. 

235U Fission

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

0.1 1 10 100 1000 10000 100000

Tim e (s)

(MeV/fission)

ORIGEN-S

ORNL

Low ell

YAYOI

Figure 2.1 Measurements of 235U fission decay heat for a fission 

pulse from ORNL, Lowell, and YAYOI experiments compared with 

calculations 
11 

Figure 2.2 Measurements of 235U fission decay heat for an 

equivalent fission pulse from Karlsruhe and Studsvik experiments 

compared with calculations 

Figure 2.3 Decay heat for 235U fission from Karlsruhe measurements 

compared with calculations for the experimental 200-s irradiation 

conditions 

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

10 100 1000 10000

Tim e (s)

(MeV / fission)

Karlsruhe

ORIGEN-S

235U Thermal Fission

235U Thermal Fission

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

1.8

10 100 1000 10000

Tim e (s)

(MeV/fission)

ORIGEN-S

Karlsruhe

Studsvik
12 

235U Thermal Fission

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

1.8

0.1 1 10 100 1000 10000 100000

Tim e (s)

(MeV/fission)

ORIGEN-S

ANS-5.1

Figure 2.4 Comparison of 235U fission decay heat for a fission 

pulse as predicted by the ANS-5.1-2005 standard with recommended 

uncertainties to the ORIGEN-S code result 

2.3.2 239Pu Fission 

Decay heat from thermal fission of 239Pu has been measured at ORNL, the University of Tokyo at the 

YAYOI reactor, and at the University of Massachusetts Lowell. All measurements were performed using 

spectroscopic techniques. As discussed previously, the YAYOI reactor measurements made by Akiyama 

et al. were performed for fast neutron irradiation. All other measurements and the ORIGEN-S simulations 

involved thermal 239Pu fission. The calculated and measured decay heat for 239Pu fission are compared in 

Figure 2.5. The figure also shows the ratio of the calculated to ORNL-measured values. The calculations 

are within the range of uncertainty of the measurements except in the cooling time range from about 300 s 

to 6000 s after fission. 

The differences between the decay heat generated for thermal neutron and fast neutron fission is expected 

to be small, < 1% (Ref. 7), and the fast experimental data (YAYOI) are combined with the thermal 

measurement data for the purpose of comparison with the calculations. 
13 

239Pu Fission

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1 10 100 1000 10000 100000

Tim e (s)

(MeV/fission)

ORIGEN-S

YAYOI

ORNL

Lowell

Figure 2.5 Measurements of 239Pu fission decay heat for a fission 

pulse ORNL, Lowell, and YAYOI experiments compared with 

calculations 

 
14 

2.3.3 241Pu Fission 

The predicted decay heat following a 241Pu thermal fission is compared with the ORNL measurements in 

Figure 2.6. No other measurements for 241Pu were found in the literature. The results show agreement 

within the assigned uncertainty of the measurements for all decay times considered. 

241Pu Thermal Fission

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

1.8

1 10 100 1000 10000 100000

Tim e (s)

(MeV/fission)

ORNL

ORIGEN-S

Figure 2.6 Comparison of 241Pu thermal fission decay heat 

for a fission pulse measured by ORNL with calculations 

2.3.4 238U Fast Fission 

Decay heat from fast fission of 238U was measured at the University of Massachusetts Lowell and at the 

YAYOI reactor facility. The predicted decay heat is compared with the measurement results in 

Figure 2.7. The Lowell measurements extend down to 0.4 s after fission, well below the range of the 

YAYOI data. Both sets of measurements are consistent with one another and are in generally good 

agreement with the predictions. The Lowell results are lower than the YAYOI results for decay times 

beyond about 5000 s. This is observed in many of the Lowell measurements for longer decay times and, 

as discussed previously, is attributed to the error in the calculated correction factor applied to account for 

noble gas loss. Thus, less weight is given to the Lowell results at longer decay times. 
15 

238U Fast Neutron Fission

0.00

0.50

1.00

1.50

2.00

2.50

0.1 1 10 100 1000 10000 100000

Tim e (s)

(MeV/fission)

ORIGEN-S

YAYOI

Low ell

Figure 2.7 Measurements of 238U decay heat for a fast fission 

pulse compared with calculations 

2.3.5 233U and 232Th Fast Fission 

The calculated decay heat following a 233U fast fission pulse event is compared with the YAYOI 

measurements in Figure 2.8. Similar results for 232Th are presented in Figure 2.9. The 232Th results 

exhibit a small underprediction in the decay heat for decay times less than about 100 s, but the 

calculations are within the two standard deviation uncertainty assigned to the measurements 

(95% confidence level). There is also an underprediction of the 232Th results beyond about 6000 s. 

The 233U results are within the assigned uncertainty of the measurements over the full range of the 

measurement data. There is a small trend to underpredict the measured 233U results in the time range near 

1000 s, as was observed for the 239Pu results. 
16 

233U Fast Neutron Fission

0.00

0.20

0.40

0.60

0.80

1.00

1.20

10 100 1000 10000 100000

Tim e (s)

(MeV/fission)

ORIGEN-S

YAYOI

Figure 2.8 Measurements of 233U decay heat (YAYOI fast reactor) 

for a fast fission pulse compared with calculations 

232Th Fast Neutron Fission

0.00

0.50

1.00

1.50

2.00

2.50

10 100 1000 10000 100000

Tim e (s)

(MeV/fission)

YAYOI

ORIGEN-S

Figure 2.9 Measurements of 232Th decay heat (YAYOI fast reactor) 

for a fast fission pulse compared with calculations 
17 

2.4 Summary of Fission Experiment Results

The predicted total energy release rate following the fission of 233U, 235U, 238U, 239Pu, 241Pu, and 232Th is 

generally seen to be within the range of experimental uncertainty over the range of the measured data. 

The independent measurements themselves also exhibit a remarkable consistency with one another. 

Nevertheless, it is noted that some earlier calorimeter measurements, not included in this study, yield 

systematically larger values of decay heat than the spectroscopic measurements that were considered for 

the present study. A separate analysis of the calorimetric and spectroscopic determinations of decay heat 

using least square methods by Tobias6

 suggests that the calorimeter results are systematically larger by 

about 1–2% over the range of the data, but that the difference is not significant at the 95% confidence 

level. 

The results presented here are generally consistent with other calculations using modern decay data.22,23 

A trend to underpredict the energy release for 239Pu fission after 103

 s is observed in most simulations. 

Investigations comparing the separate contributions of gamma-ray energy and beta energy (available with 

the spectroscopic measurements) attribute the discrepancy to an error in the gamma contribution likely 

due to either the evaluated decay scheme or the gamma energy per decay.23 Regarding the current level 

of discrepancy between measurements and predictions, it has been suggested by Dickens,24 that further 

experiments at the present measurement uncertainty level are unlikely to resolve the small differences 

presently observed between predictions and measurements and that more effort is likely needed in 

research into decay schemes for the short-lived fission products. Work to improve the fundamental decay 

data used in these calculations has been initiated through the Organization for Economic Cooperation and 

Development, Nuclear Energy Agency (OECD/NEA) Working Party on International Nuclear Data 

Evaluation Cooperation (WPEC) Fission Product Decay Heat Subgroup 25.25

A quantitative uncertainty analysis of the decay heat benchmark results for the fission experiments was 

beyond the scope of this report. Additional work in this area will be performed in the context of future 

development of the ANS-5.1 Standard. Nevertheless the present results demonstrate that the ORIGEN-S 

calculations are generally within the uncertainty range of the experimental data measured by 

spectroscopic techniques for most of the fissionable nuclides of importance to short-term decay heat in 

commercial and research reactor fuels. 
18 
19 

3. ASSEMBLY MEASUREMENTS 

Decay heat measurements of full-length spent fuel assemblies performed at two separate U.S. facilities 

have been reported: (1) using a pool calorimeter located at the General Electric Morris Operations 

(GE-Morris) spent fuel storage facility in Morris, Illinois, and (2) using a boil-off calorimeter operated at 

the Engine Maintenance Assembly and Disassembly (EMAD) facility located on the Nevada Test Site of 

Hanford Engineering Development Laboratory (HEDL). Decay heat was measured for fuel assemblies 

discharged from San Onofre 1, Point Beach, and Turkey Point PWRs, and from the Dresden, Cooper and 

Monticello BWRs. In total, measurements for 65 different assemblies at the GE-Morris and HEDL 

calorimeters have been reported, with multiple measurements performed on many of these assemblies. 

More recent calorimeter measurements of fuel assemblies were performed at the Swedish Central Interim 

Storage Facility for Spent Fuel (CLAB), located in Oskarshamn, Sweden. This project, managed by 

SKB, has established a large database of spent fuel measurements and will continue to perform 

measurements in the future to ensure that changes in the decay heat are accurately simulated over time. 

The measurements performed to date include 39 BWR assemblies and 34 PWR assemblies from 

Barsebäck, Ringhals, Forsmark, and Oskarshamn nuclear power plants in Sweden. The following 

sections present a brief summary of the assembly measurements and describe the calorimeters operated at 

each of the three measurement facilities mentioned above. 

3.1 Facility Descriptions

3.1.1 Hanford Engineering Development Laboratory 

The Hanford Engineering Development Laboratory fuel assembly calorimeter26 operated at the Engine 

Maintenance Assembly and Disassembly facility on the Nevada Test Site was designed by Pacific 

Northwest Laboratory to measure intact PWR and BWR spent fuel assemblies with decay heat rates in the 

range of 0.1 to 2.5 kW. The accuracy of the measurements was estimated to be ± 5% for decay heat rates 

greater than about 1000 W, and ± 10 % at 100 W. The calorimeter is comprised of a double-walled 

stainless steel vessel approximately 5.49 m in length. The inner wall of the vessel supports a lead shield 

that acts to capture and account for the leakage of gamma decay energy from the calorimeter. The 

calorimeter system is illustrated in Figure 3.1. 

The HEDL calorimeter measures decay heat by evaluating differential steam condensate collection rates, 

and is referred to as a boil-off type calorimeter. The calorimeter is first filled with water and brought to 

boiling conditions using a precision internal heater. When equilibrium steam condensate collection rate is 

established, the spent fuel assembly is placed within the calorimeter vessel and a boiling equilibrium 

condition is then reestablished with the combined heat inputs of the assembly and the vessel heater. 

The decay heat from the assembly is then determined directly from the differential condensate collection 

rate of the initial reference state and the final equilibrium state. 

Decay heat measurements using the HEDL calorimeter were reported for four Westinghouse design 

15 × 15 fuel assemblies from the Turkey Point reactor.27,28 Cooling times for these assemblies were 

approximately 2.5 years, and the discharge burnups were all between 26 and 28 GWd/MTU. 
20 

Figure 3.1 HEDL Spent Fuel Calorimeter 

(Source: Schmittroth et al., Ref. 27) 
21 

3.1.2 GE-Morris Operation 

The majority of spent fuel assembly decay heat measurements performed in the United States were made 

at a calorimeter operated at the GE-Morris Operation facility.29 The calorimeter, located underwater in 

the spent fuel storage bay, consists of two concentric carbon steel pipes separated by urethane foam 

insulation to reduce heat loss to the pool. The calorimeter, shown schematically in Figures 3.2 and 3.3, is 

4.6 m long (excluding the detachable head assembly). The calorimeter contains inserts for PWR and 

BWR fuels that maintain the assemblies in a centered position. Fuel assemblies are loaded into the 

calorimeter through the vessel head in a procedure similar to loading a fuel transport cask, then the head 

is installed, and a leak check is performed. 

The principal method used to determine the thermal output of the assembly, referred to as static mode, is 

to measure the rate of the water temperature increase inside the vessel. The relationship between the rate 

of temperature rise and the thermal output is determined by calibration using an electric heater assembly. 

A second method, referred to as recirculation mode, involves circulating water through the vessel and 

determining the thermal output from the temperature change of the water flowing through the vessel and 

the rate of water flow. This second method was used early in the program but was not widely employed 

because it was found to be less reliable than the static mode of operation. The temperature rise in the 

vessel is measured using a series of platinum resistance temperature detectors (RTDs) mounted on the 

assembly insert, vessel wall, and at the inlet and outlet valves. The decay energy escaping the calorimeter 

as gamma radiation is measured using gamma ray detectors located at several elevations along the 

calorimeter’s surface. 

The systematic error associated with the static-mode calorimeter measurements was estimated to be about 

± 2% for a thermal output of about 700 W, increasing to ± 4% in the 200-W range. The random error 

associated with the repeatability of measurements for the same assembly was determined in test runs 

involving 14 measurements of Cooper assembly CZ205 over an 8-month period30 and yielded a random 

error of about ± 4.3%. The decay heat output of assembly CZ205 at the time of the test measurements 

was in the 300-W range. 
22 

Figure 3.2 Schematic of GE-Morris calorimeter system components 

(Source: McKinnon et al., Ref. 30) 
23 

Figure 3.3 Isomeric view of GE-Morris calorimeter 

(Source: Judson et al., Ref. 29) 
24 

3.1.3 Swedish Central Interim Storage Facility for Spent Fuel 

The Swedish Nuclear Fuel and Waste Management Co., SKB, operates the Central Interim Storage 

Facility for Spent Fuel, known as CLAB, on the east coast of Sweden near the site of the Oskarshamn 

nuclear power plant. To support the characterization of the spent fuel inventory at CLAB, SKB has 

constructed a spent fuel assembly calorimeter and performed measurements for use in validating 

computer code predictions of decay heat. Measurements on selected fuel assemblies are to be repeated at 

intervals in the future, and additional assemblies may be added as required to validate computer code 

predictions over the range of fuel inventory at CLAB.5

The design of the calorimeter at CLAB is based on the calorimeter design used at GE Morris Operation. 

The calorimeter is 4.9 m long and composed of two concentric pipes separated by polyurethane foam to 

provide insulation between the interior vessel and the pool water. There is a fixed insert for PWR fuel 

assemblies and a removable insert for BWR assemblies used to maintain the fuel assembly in a centered 

and vertical position in the calorimeter. Temperatures within the calorimeter are measured by 16 PT-100 

type sensors: eight sensors in the water inside the vessel; two sensors on the inside and outside surfaces 

of the calorimeter; and two sensors outside the calorimeter to measure the pool water temperature. There 

are five gamma monitors positioned to measure the radial distribution of gamma radiation that escapes the 

calorimeter, and this information is used to correct the calorimeter data for gamma energy loss. 

The calorimeter is generally operated in static mode using the rate of water temperature increase to 

establish the thermal output of the assemblies. The calorimeter can also be operated using recirculation 

mode (measurement of temperature difference between inlet and outlet) and in an equilibrium mode 

whereby the equilibrium temperature is reached without circulation. However, both of these latter 

methods require substantially longer measurement times that the static mode. In static mode the decay 

heat is evaluated by comparing the temperature increase to a calibration curve established using an 

electric heated assembly prior to each measurement campaign. The calibration curves are made for a 

series of power levels that allow the rate of temperature rise to be determined as a function of assembly 

decay heat. The decay heat established by the measurements is then corrected for energy loss from 

gamma rays escaping from the calorimeter using data from the external gamma detectors. The 

calorimeter, located in the spent fuel pool, is shown in Figure 3.4. 

The design target accuracy for the calorimeter was about ± 2% at the 95% confidence level. The random 

error associated with the measurements, based on the final design of the calorimeter, was estimated by 

SKB as a function of decay heat rate: for a low thermal output of about 50 W, the estimated 95% error is 

± 4.5 W, or 9%. For a thermal output of about 300 W, the error is about ± 8 W or 2.7%, decreasing 

relatively to ± 15.5 W or 1.7% at 900 W of generated heat. The actual random error was made using 

eight repeat measurements of Ringhals-1 8 × 8 assembly 6432 over a 9-month period. The standard 

deviation of the data is about 0.9% for the average decay heat value of 185 W over the measurement 

period. This yields a 95% confidence error of less than 2% which is less than the estimated error. 

Evaluation of other repeat measurements for six of the BWR assemblies and four of the PWR assemblies 

yielded a standard deviation for this group of assemblies to be 3.94 W. For the BWR assemblies only, the 

value was 1.4 W, and for the PWR assemblies it was 5.8 W. These results are consistent with the random 

error estimated by SKB. The uncertainty analysis described includes only random sources of uncertainty 

and does not address potential sources of systematic bias. 
25 

Figure 3.4 Spent fuel assembly calorimeter at the Swedish CLAB facility located at 

poolside and raised to show the assembly loading hatch above the water level 

3.2 Summary of Measurements

A comprehensive summary of measurements performed in the U.S. has been compiled by Roddy and 

Mailen.31 These measurements, performed at the HEDL and GE-Morris Operation facilities, include 

spent fuel from San Onofre 1, Point Beach 2, Turkey Point 3, Dresden, Cooper and Monticello nuclear 

plants. The measurements of the Monticello assemblies, performed at the GE-Morris facility in 1985, 

represent the last measurements performed in the United States. No measurements have been reported 

after this date, and the calorimeters have since been decommissioned. 

Table 3.1 summarizes general information on the fuel assemblies measured at each facility, including 

assembly design, range of fuel enrichment, and maximum assembly burnup. The number of 

measurements is actually greater than the number of different assemblies indicated in the table because 

many of the assemblies were remeasured at different decay times. 

Several rebuilt fuel assemblies (reconfiguration or replacement of the fuel rods in the assembly) were 

measured at the CLAB facility. However, because the details of the rebuilts were not available at this 

time, these assemblies (not included in the data shown in Table 3.1) were excluded in the present analysis. 
26 

Table 3.1 Summary of evaluated fuel assembly decay heat measurements 

Measurement 

facility Reactor Fuel design Enrichment(s) 

(wt % 235U) 

Number of 

assemblies 

measured 

Maximum 

burnup 

(MWd/MTU) 

 San Onofre 1 W 14 × 14Sa 3.865 – 4.005 8 32,363 

 Point Beach 2 W 14 × 14 3.397 6 39,384 

GE-Morris Dresden 2 GE 7 × 7 2.128 1 5,280 

 Cooper GE 7 × 7 1.1, 2.50 54 28,048 

 Monticello GE 7 × 7 2.25 6 20,189 

 

HEDL (EMAD ) Turkey Point W 15 × 15 2.557 4 28,588 

 

 Barsebäck 1 8 × 8 2.922 – 2.953 2 41,127 

 Barsebäck 2 8 × 8 3.154 1 40,010 

 Forsmark 1 8 × 8 2.090 – 2.970 2 34,193 

 Forsmark 1 9 × 9 2.938 3 37,896 

Swedish CLAB Forsmark 2 8 × 8 2.095 1 19,944 

 Forsmark 2 SVEA-64 2.850 – 2.920 3 32,837 

 Forsmark 3 SVEA-100 2.770 2 31,275 

 Oskarshamn 2 8 × 8 2.201 – 2.875 7 34,893 

 Oskarshamn 2 SVEA-64 2.902 1 46,648 

 Oskarshamn 3 8 × 8 2.577 1 35,619 

 Oskarshamn 3 SVEA-100 2.711 2 40,363 

 Ringhals 1 8 × 8 2.640, 2.911 9 44,861 

 Ringhals 2 15 × 15 3.095 – 3.252 16 50,962 

 Ringhals 3 17 × 17 2.100 – 3.404 14 47,308 

a

 Stainless steel cladding. 
27 

4. FUEL ASSEMBLY DATA 

This section presents fuel design information and reactor operating data in sufficient detail to allow an 

independent evaluation of the decay heat measurements analyzed in this report. Additional and more 

detailed information is available from the primary references cited. The fuel design data are used to 

prepare physics models of the assemblies (presented in Sect. 5) with modules of the SCALE 5 code 

system from which assembly cross-section libraries are generated for the ORIGEN-S fuel depletion 

analyses. The measurement data are summarized in Appendix A, and the analysis of the measurements 

are presented in Sect. 6 along with the results computed using the ORIGEN-S code. 

4.1 Assembly Designs—U.S. Measurements

The description of the measured assemblies from the San Onofre, Turkey Point, Point Beach, Cooper, 

Dresden, and Monticello reactors was obtained from a number of different sources. A common source of 

information for all U.S.-origin fuel was the data compiled by the Energy Information Administration 

(EIA) of the U.S. Department of Energy (DOE) in Nuclear Fuel Data survey Form RW-859.32 The 

information collected on Form RW-859 contains data on every fuel assembly irradiated in commercial 

nuclear reactors operating in the United States. The data are considered in the design and operation of the 

equipment and facilities that will be used by DOE for the future acceptance, transport, and disposal of all 

domestic spent fuel. 

The specific fuel data from RW-859 used in this study included the assembly type, version (EIA code), 

initial uranium mass, enrichment, discharge cycle, and discharge burnup as reported by the utilities. 

Table 4.1 gives a summary of all U.S. fuel measurements evaluated in this study. 

4.1.1 San Onofre 1 Assemblies 

Decay heat measurements performed at the GE-Morris facility are reported for eight spent fuel assemblies 

from the San Onofre Unit 1 PWR reactor in report NEDG-24922-3 (Ref. 29). Fuel information and 

detailed operating history data are reported separately by Schmittroth in report HEDL-TME 83-32 

(Ref. 28). The assemblies were all Westinghouse 14 × 14 design. The information in RW-859 identifies 

the assembly version as EIA code XSO14W. The design specifications for this assembly were obtained 

from a DOE report on the characteristics of potential repository waste DOE/RW-0184 (Ref. 33). The 

assembly design data used for the calculations are given in Table 4.2. The fuel rod configuration of the 

assembly, based on models used in Ref. 34, is illustrated in Figure 4.1. 

The San Onofre fuel assemblies used Type 304 stainless steel cladding. Activated cobalt in the steel 

cladding is an important component of the decay heat during the timeframe of the measurements and must 

be taken into account. The mass of the stainless steel guide tubes is given as 7.2 kg per assembly in 

DOE/RW-0184, Vol. 3 (Ref. 33). The mass of the stainless steel cladding for the 180 fuel rods is 

calculated as 59.1 kg, yielding a total steel mass of 66.3 kg per assembly. This value is significantly less 

than the 150-kg mass used in HEDL TME-83-32 (Ref. 28). This difference may be a result of the 

previous calculations including components located away from the fueled sections of the assembly. 

However, since the flux decreases dramatically with distance from the fuel region, it is not appropriate to 

include these components at their actual mass. Because the fraction of activation occurring in these 

regions is small compared with that in the fuel region, only the amounts of steel associated with the 

cladding and guide tube structures were included in the present calculations. A cobalt impurity level of 

800 ppm in stainless steel was assumed, a value obtained from DOE/RW-0184. The other constituents of 

the stainless steel are listed in Table 4.3, although cobalt is the dominant impurity in terms of the 

contribution to decay heat. Inconel-718, also present in the assembly as spacers and a hold-down spring, 
28 

was not included because of its relatively low importance as compared with the cladding and guide tubes. 

The total amount of cobalt present in the assembly cladding and guide tubes was estimated to be 53 g. 

The fuel assembly irradiation data for the measured fuel assemblies designated C-01, C-16, C-19, C-20, 

D-01, D-46, E-18 and F-04 were obtained from Schmittroth28 and are listed here in Table 4.4. The 

assembly irradiation data in Table 4.4 are consistent with the discharge burnup from RW-859. 

Table 4.1 Summary of fuel data for U.S. assembly measurements 

No. Reactor Assembly 

type 

Assembly 

ID 

Enrichment 

(wt % 235U) 

Uranium 

(kg) 

Discharge 

burnup 

(MWd/MTU) 

Measurement 

facility 

1 Cooper GE 7 × 7 CZ102 1.090 195.48 11,667 GE-Morris 

2 Cooper GE 7 × 7 CZ147 2.500 190.31 26,709 GE-Morris 

3 Cooper GE 7 × 7 CZ148 2.500 190.22 26,310 GE-Morris 

4 Cooper GE 7 × 7 CZ182 2.500 190.09 26,824 GE-Morris 

5 Cooper GE 7 × 7 CZ195 2.500 190.68 26,392 GE-Morris 

6 Cooper GE 7 × 7 CZ205 2.500 190.72 25,344 GE-Morris 

7 Cooper GE 7 × 7 CZ209 2.500 190.38 25,383 GE-Morris 

8 Cooper GE 7 × 7 CZ211 2.500 190.82 26,668 GE-Morris 

9 Cooper GE 7 × 7 CZ222 2.500 190.90 26,692 GE-Morris 

10 Cooper GE 7 × 7 CZ225 2.500 190.51 25,796 GE-Morris 

11 Cooper GE 7 × 7 CZ239 2.500 189.57 27,246 GE-Morris 

12 Cooper GE 7 × 7 CZ246 2.500 189.81 27,363 GE-Morris 

13 Cooper GE 7 × 7 CZ259 2.500 190.20 26,466 GE-Morris 

14 Cooper GE 7 × 7 CZ264 2.500 190.89 26,496 GE-Morris 

15 Cooper GE 7 × 7 CZ277 2.500 189.49 26,748 GE-Morris 

16 Cooper GE 7 × 7 CZ286 2.500 189.95 27,141 GE-Morris 

17 Cooper GE 7 × 7 CZ296 2.500 190.50 26,388 GE-Morris 

18 Cooper GE 7 × 7 CZ302 2.500 190.00 26,594 GE-Morris 

19 Cooper GE 7 × 7 CZ308 2.500 189.78 25,815 GE-Morris 

20 Cooper GE 7 × 7 CZ311 2.500 189.91 27,392 GE-Morris 

21 Cooper GE 7 × 7 CZ315 2.500 189.96 26,881 GE-Morris 

22 Cooper GE 7 × 7 CZ318 2.500 189.32 26,568 GE-Morris 

23 Cooper GE 7 × 7 CZ331 2.500 190.36 21,332 GE-Morris 

24 Cooper GE 7 × 7 CZ337 2.500 189.90 26,720 GE-Morris 

25 Cooper GE 7 × 7 CZ342 2.500 190.16 27,066 GE-Morris 

26 Cooper GE 7 × 7 CZ346 2.500 190.23 28,048 GE-Morris 

27 Cooper GE 7 × 7 CZ348 2.500 190.38 27,481 GE-Morris 

28 Cooper GE 7 × 7 CZ351 2.500 190.02 25,753 GE-Morris 

29 Cooper GE 7 × 7 CZ355 2.500 190.60 25,419 GE-Morris 

30 Cooper GE 7 × 7 CZ357 2.500 190.19 27,140 GE-Morris 

31 Cooper GE 7 × 7 CZ369 2.500 190.20 26,576 GE-Morris 

32 Cooper GE 7 × 7 CZ370 2.500 190.23 26,342 GE-Morris 

33 Cooper GE 7 × 7 CZ372 2.500 190.01 25,848 GE-Morris 

34 Cooper GE 7 × 7 CZ379 2.500 190.18 25,925 GE-Morris 

35 Cooper GE 7 × 7 CZ398 2.500 189.83 27,478 GE-Morris 

36 Cooper GE 7 × 7 CZ415 2.500 189.72 25,863 GE-Morris 

37 Cooper GE 7 × 7 CZ416 2.500 189.43 27,461 GE-Morris 

38 Cooper GE 7 × 7 CZ429 2.500 190.07 27,641 GE-Morris 
29 

Table 4.1 Summary of fuel data for U.S. assembly measurements (continued) 

No. Reactor Assembly 

type 

Assembly 

ID 

Enrichment 

(wt % 235U) 

Uranium 

(kg) 

Discharge 

burnup 

(MWd/MTU) 

Measurement 

facility 

39 Cooper GE 7 × 7 CZ430 2.500 189.93 26,825 GE-Morris 

40 Cooper GE 7 × 7 CZ433 2.500 190.02 25,977 GE-Morris 

41 Cooper GE 7 × 7 CZ460 2.500 190.18 26,512 GE-Morris 

42 Cooper GE 7 × 7 CZ466 2.500 189.86 26,077 GE-Morris 

43 Cooper GE 7 × 7 CZ468 2.500 189.78 26,757 GE-Morris 

44 Cooper GE 7 × 7 CZ472 2.500 190.12 25,957 GE-Morris 

45 Cooper GE 7 × 7 CZ473 2.500 189.76 26,519 GE-Morris 

46 Cooper GE 7 × 7 CZ498 2.500 189.69 26,482 GE-Morris 

47 Cooper GE 7 × 7 CZ508 2.500 190.68 26,357 GE-Morris 

48 Cooper GE 7 × 7 CZ515 2.500 190.48 25,737 GE-Morris 

49 Cooper GE 7 × 7 CZ526 2.500 190.54 27,596 GE-Morris 

50 Cooper GE 7 × 7 CZ528 2.500 190.81 25,715 GE-Morris 

51 Cooper GE 7 × 7 CZ531 2.500 189.90 26,699 GE-Morris 

52 Cooper GE 7 × 7 CZ536 2.500 190.17 26,715 GE-Morris 

53 Cooper GE 7 × 7 CZ542 2.500 189.99 26,691 GE-Morris 

54 Cooper GE 7 × 7 CZ545 2.500 190.47 26,668 GE-Morris 

55 Dresden GE 7 × 7 DN212 2.130 194.70 5,279 GE-Morris 

56 Monticello GE 7 × 7 MT116 2.250 193.53 18,039 GE-Morris 

57 Monticello GE 7 × 7 MT123 2.250 193.53 13,027 GE-Morris 

58 Monticello GE 7 × 7 MT133 2.250 193.53 20,994 GE-Morris 

59 Monticello GE 7 × 7 MT190 2.250 193.53 15,143 GE-Morris 

60 Monticello GE 7 × 7 MT228 2.250 193.53 12,123 GE-Morris 

61 Monticello GE 7 × 7 MT264 2.250 193.53 9,160 GE-Morris 

62 Point Beach 2 W 14 × 14 C-52 3.397 386.54 31,914 GE-Morris 

63 Point Beach 2 W 14 × 14 C-56 3.397 386.80 38,917 GE-Morris 

64 Point Beach 2 W 14 × 14 C-64 3.397 386.63 39,384 GE-Morris 

65 Point Beach 2 W 14 × 14 C-66 3.397 386.54 35,433 GE-Morris 

66 Point Beach 2 W 14 × 14 C-67 3.397 386.45 38,946 GE-Morris 

67 Point Beach 2 W 14 × 14 C-68 3.397 386.36 37,059 GE-Morris 

68 San Onofre1 W 14 × 14a

 C-01 3.865 361.72 26,540 GE-Morris 

69 San Onofre 1 W 14 × 14 a C-16 3.865 361.72 28,462 GE-Morris 

70 San Onofre 1 W 14 × 14 a C-19 3.865 361.72 30,426 GE-Morris 

71 San Onofre 1 W 14 × 14 a C-20 3.865 361.72 32,363 GE-Morris 

72 San Onofre 1 W 14 × 14 a D-01 4.005 363.64 31,393 GE-Morris 

73 San Onofre 1 W 14 × 14 a D-46 4.005 363.64 32,318 GE-Morris 

74 San Onofre 1 W 14 × 14 a E-18 4.005 363.98 32,357 GE-Morris 

75 San Onofre 1 W 14 × 14 a F-04 3.996 372.32 30,429 GE-Morris 

76 Turkey Point 3 W 15 × 15 B-43 2.559 447.79 25,595 Hanford 

77 Turkey Point 3 W 15 × 15 D-15 2.557 456.12 28,430 Hanford 

78 Turkey Point 3 W 15 × 15 D-22 2.557 458.00 26,485 Hanford 

79 Turkey Point 3 W 15 × 15 D-34 2.557 455.24 27,863 Hanford 

a

 Stainless steel cladding. 
30 

Table 4.2 San Onofre Unit 1 reactor and fuel assembly design information 

Parameter Data 

Assembly and reactor data 

Designer Westinghouse Electric 

Lattice geometry 14 × 14 

EIA assembly codea

 XSO14W 

Active fuel rod length (cm) 304.8 

Uranium mass per assembly (kg) 366 (nominal) 

Rod pitch (cm) 1.4122 

Number of fuel rods 180 

Number of instrument tubes 1 

Number of guide tubes 15 

Moderator pressure (psia) 2100 

Assembly pitch (cm) 19.94 

Inlet temperature (K) 562.4 

Outlet temperature (K) 586.9 

Average moderator temperature (K) 576.5 

Moderator density (g/cm3

) 0.7179 

Average soluble boron level (ppm) 500 

 

Fuel rod data 

Fuel material type UO2

Fuel pellet density (% TD) 93 (10.19 g/cm3) 

Stack density (g/cm3

) 10.037 

Fuel pellet diameter (cm) 0.9741 

Fuel rod outside diameter (cm) 1.0719 

Effective fuel temperature (K) 810 

Clad material Stainless steel – Type 304 

Clad thickness (cm) 0.0419 

Inner clad diameter (cm) 0.9881 

Average clad temperature (K) 615 

 

Guide tube data 

Guide tube material Stainless steel – Type 304 

Inside diameter (cm) 1.2979 

Outside diameter (cm) 1.3589 

a

 Assigned by Department of Energy Information Administration (EIA) 

Form RW-859. 
31 

Figure 4.1 Layout of the San Onofre Westinghouse 14 × 14 assembly 
32 

Table 4.3 Stainless steel composition data for San Onofre fuel assemblies 

Element parts per million 

(ppm)a

grams per 

assemblyb kg per MTUc

Carbon 800 53 0.146 

Nitrogen 1,300 86.1 0.238 

Silicon 10,000 662.6 1.830 

Phosphorous 450 29.8 0.082 

Sulfur 300 19.9 0.055 

Chromium 190,000 12,589 34.77 

Manganese 20,000 1325 3.66 

Iron 697,150 46,193 127.6 

Cobalt 800 53.0 0.146 

Nickel 89,200 5910 16.33 

Niobium 100 6.6 0.018 

a From Table 2.7.3 of reference DOE/RW-0184-R1, Vol. 1 (Ref. 33). 

b

 Quantities based on nominal 66.3 kg stainless steel per assembly. 

c

 Metric Ton Uranium (106 g); values assume nominal 363 kg U per assembly. 

Table 4.4 San Onofre Unit 1 reactor and assembly irradiation data

Operational data Cycle 1 Cycle 2 Cycle 3 Cycle 4 Cycle 5 Cycle 6

Startup date 1/27/68 11/20/70 2/24/72 6/30/73 4/23/75 4/04/77

Shutdown date 10/2/70 1/1/72 6/2/73 3/14/75 9/30/76 9/15/78

Operating days 979 407 464 622 526 529

EFPD daysa 672 384 436 490 522 530

Downtime (days) 49 54 28 40 186 −

Assembly ID Assembly burnup (MWd/MTU) b

C-01 7,689 16,205 26,540 

C-16 9,077 18,452 28,462 

C-19 11,750 21,372 30,426 

C-20 14,418 23,410 32,363 

D-01 9,738 20,592 31,393 

D-46 10,029 21,055 32,318 

E-18 8,068 19,470 32,357 

F-04 6,883 19,733 30,429

a

 EFPD = effective full-power days. 

b

 Cumulative burnup by cycle. 
33 

4.1.2 Turkey Point 3 Assemblies 

Decay heat measurements were performed on five spent fuel assemblies from the Turkey Point Unit 3 

PWR, operated by Florida Power and Light Company, using the boil-off calorimeter at the HEDL facility. 

The assemblies, B-43, D-04, D-15, D-22 and D-34, all have an EIA version code W1515WL, 

corresponding to a Westinghouse 15 × 15 LOPAR design, as shown in Figure 4.2. Decay heat 

measurements for these assemblies were previously evaluated27,28 against code predictions and more 

recently used to support the development of technical guidance3

 for the NRC decay heat regulatory 

guidance RG 3.54 (Ref. 4). 

The Turkey Point fuel and assembly design specifications are listed in Table 4.5. Each assembly contains 

18.84 kg of stainless steel Type 304, 4.65 kg of Inconel-718, and 110.0 kg of Zircaloy-4. The reactor 

cycle information, assembly power, and discharge burnup values were taken from the values cited in the 

report HEDL-TME 83-32 (Ref. 28). The discharge burnup values differed by several percent (4% in the 

case of assembly B-43) as compared with the values reported in RW-859. The values from report 

HEDL-TME 83-32 are similar to those reported by Florida Power and Light Co. and used in the analyses 

performed in Ref. 3. The reason for the burnup discrepancy for these Turkey Point assemblies, 

particularly B-43, is unclear. Such discrepancies were not observed for any other assemblies used in this 

study. The decay times were determined in accordance with the actual date of the measurement and a 

Cycle 4 discharge date from RW-859 that was slightly different than that used in earlier calculations. 

The reactor operating history data and assembly burnup are listed in Table 4.6. Structural materials 

included for activation are listed in Table 4.7. 

A measurement of assembly D-04 was rejected in this study. HEDL-TME 83-32 reports that the thermal 

equilibrium had not been established for the D-04 measurement, thus potentially biasing the measured 

value low. In addition, report TC-1759 (Ref. 27) indicated that four fuel rods had been removed from this 

assembly at the time of measurements. Thus, there is additional uncertainty in the results for this 

particular assembly. 
34 

Figure 4.2 Layout of the Turkey Point Reactor Westinghouse 15 × 15 assembly 
35 

Table 4.5 Turkey Point 3 reactor and fuel assembly design information 

Parameter Data 

Assembly and reactor data 

Designer Westinghouse Electric 

Lattice geometry 15 × 15 

EIA assembly code W1515WL 

Version LOPAR 

Active fuel rod length (cm) 365.8 

Uranium mass per assembly (kg) 456 (nominal) 

Rod pitch (cm) 1.430 

Number of fuel rods 204 

Number of instrument tubes 1 

Number of guide tubes 20 

Moderator pressure (psia) 2100 

Assembly pitch (cm) 21.50 

Inlet temperature (K) 558.7 

Outlet temperature (K) 611.0 

Average moderator temperature (K) 570 

Moderator density (g/cm3

) 0.7311 

Average soluble boron level (ppm) 450 

 

Fuel rod data 

Fuel material type UO2

Fuel pellet density (% TD) 93 

Stack density (g/cm3

) 10.19 

Fuel pellet diameter (cm) 0.9296 

Fuel rod outside diameter (cm) 1.0719 

Effective fuel temperature (K) 922 

Clad material Zircaloy-4 

Clad thickness (cm) 0.0618 

Inner clad diameter (cm) 0.9484 

Average clad temperature (K) 595 

 

Guide tube data 

Guide tube material Zircaloy-4 

Inside diameter (cm) 1.3000 

Outside diameter (cm) 1.3868 
36 

Table 4.6 Turkey Point reactor and assembly irradiation data 

Operational data Cycle 1 Cycle 2 Cycle 3 Cycle 4

Startup date 10/20/1972 12/16/1974 12/23/1975 1/16/1977

Shutdown date 10/4/1974 10/26/1975 11/15/1976 11/24/1977

Operating days 714 314 328 312

Downtime (days) 73 58 62 --

Assembly ID Assembly Burnup (MWd/MTU) 

B-43 16,493 25,595 

D-15 9,574 19,422 28,430

D-22 10,030 19,081 26,485

D-34 9,571 18,992 27,863

Table 4.7 Activated structural components in 

Turkey Point and Point Beach assemblies 

kilograms per assembly Element grams per kg Ua

14 × 14 assemblyb 15 × 15 assemblyc

Oxygen 135 52.0 62.0 

Chromium 5.9 2.3 2.7 

Manganese 0.33 0.13 0.15 

Iron 12.9 5.0 5.9 

Cobalt 0.075 0.029 0.034 

Nickel 9.9 3.8 4.5 

Zirconium 221 85.0 101 

Niobium 0.71 0.27 0.32 

Tin 3.6 1.4 1.6 

a Source: NUREG/CR-5625 (Ref. 3). 

b

 Assumed mass of 389 kg U per assembly. 

c

 Assumed mass of 456 kg U per assembly. 
37 

4.1.3 Point Beach Assemblies 

Decay heat measurements were performed at the GE-Morris Operation facility for six spent fuel 

assemblies from the Point Beach Unit 2 PWR, operated by Wisconsin Electric Power Co. (WEPCO). 

The results are identified in earlier studies as the WEPCO measurements. 28,29 The Point Beach reactor 

assemblies were also previously evaluated in support of technical guidance for the NRC decay heat 

regulatory guidance RG 3.54 issued in 1999 (Ref. 4). The Point Beach reactor assemblies C-52, C-56, 

C-64, C-66, C-67, and C-68 have an EIA designation code W1414WL, corresponding to the 

Westinghouse 14 × 14 LOPAR design. The LOPAR design is similar to the San Onofre assembly design 

but has 179 fuel rods (instead of 180) and has a longer active fuel length and thus larger uranium mass. 

The fuel and assembly design specifications are listed in Table 4.8. Activated structural assembly 

constituents are listed in Table 4.7. The fuel rod layout in the assembly is shown in Figure 4.3. 

The Point Beach assemblies resided in the core during Cycles 1, 2, and 3 and achieved burnup values 

from 31,914 to 39,384 MWd/MTU. The measurements were performed approximately 4.5 years after 

discharge. The enrichment of all measured assemblies was 3.397 wt % 235U. The assembly burnup and 

reactor operating data are listed in Table 4.9. The reactor operated at near 20% capacity for about the first 

40 weeks of Cycle 1. Ascent to full power occurred near the beginning of May 1973. These two phases 

of operation during Cycle 1 are represented separately as two steps (Cycle 1A and 1B) to more accurately 

represent the low-power startup operation in the simulations. The assembly burnups for these two phases 

were derived from the operating times and core-average burnup values from Ref. 28. 
38 

Table 4.8 Point Beach reactor and fuel assembly design information 

Parameter Data 

Assembly and reactor data 

Designer Westinghouse 

Lattice geometry 14 × 14 

EIA assembly code W1414WL 

Version LOPAR 

Active fuel rod length (cm) 365.8 

Uranium mass per assembly (kg) 389 (nominal) 

Rod pitch (cm) 1.412 

Number of fuel rods 179 

Number of instrument tubes 1 

Number of guide tubes 16 

Moderator pressure (psia) 2000 

Assembly pitch (cm) 19.82 

Inlet temperature (K) 558.0 

Outlet temperature (K) 589.0 

Average moderator temperature (K) 579 

Moderator density (g/cm3

) 0.7115 

Average soluble boron level (ppm) 550 

 

Fuel rod data 

Fuel material type UO2

Fuel pellet density (% TD) 93 

Stack density (g/cm3) 9.467 

Fuel pellet diameter (cm) 0.9290 

Fuel rod outside diameter (cm) 1.0719 

Effective fuel temperature (K) 811 

Clad material Zircaloy-4 

Clad thickness (cm) 0.0618 

Inner clad diameter (cm) 0.9484 

Average clad temperature (K) 620 

 

Guide tube data 

Guide tube material Zircaloy-4 

Inside diameter (cm) 1.2827 

Outside diameter (cm) 1.3691 
39 

Figure 4.3 Layout of the Point Beach Reactor Westinghouse 14 × 14 (LOPAR) assembly 
40 

Table 4.9 Point Beach reactor operating history data 

Operational data Cycle 1A a Cycle 1B Cycle 2 Cycle 3 

Startup date 8/1/1972 5/1/1973 12/20/1974 3/29/1976 

Shutdown date 5/1/1973 10/16/1974 2/26/1976 3/3/1977 

Operating days 273 533 433 339 

Downtime (days) 0 65 32 -- 

Assembly ID Assembly Burnup (MWd/MTU)b

C-52 677 10,801 23,117 31,914 

C-56 1,033 16,475 29,356 38,917 

C-64 1,061 16,920 29,764 39,384 

C-66 732 11,668 24,924 35,433 

C-67 1,041 16,600 29,401 38,946 

C-68 817 13,034 26,942 37,059 

a Cycle 1 is divided to present initial phase of low-power operation.

b

 Cumulative burnup by cycle. 
41 

4.1.4 Cooper, Dresden, and Monticello Assemblies 

The largest number of decay heat measurements performed for any single assembly design type were for 

General Electric (GE) 7 × 7 assemblies from the Cooper, Dresden, and Monticello reactors. All of the 

GE 7 × 7 measurements were performed at the GE-Morris facility. Two measurements were made for a 

single Dresden reactor assembly, 13 measurements for Monticello reactor assemblies, and 

81 measurements for assemblies from the Cooper reactor. The RW-859 data indicate that the assemblies 

from Cooper included two versions of the GE 7 × 7 assembly; one having an EIA code G4607G2, the 

other having code G4607G3BD. These codes correspond to GE assembly versions GE-2 and GE-3B, 

respectively. The assemblies from the Dresden and Monticello reactors – EIA code G2307G2B – were 

both versions GE-2B. 

The fuel assembly design specification for assembly versions GE-2, GE-2B, and GE-3B were obtained 

from DOE/RW-0184-R1 (Ref. 33) and are listed in Table 4.10. All assemblies have similar dimensions. 

However, the GE-3 fuel design has increased cladding thickness compared with the GE-2 design to 

reduce fuel failure caused by clad hydriding. The increase in cladding thickness decreased the pellet 

diameter and decreased the mass of uranium per assembly. The GE 7 × 7 assembly configuration is 

illustrated in Figures 4.4 and 4.5. 

Fuel and moderator temperatures and moderator and channel water densities were not reported. The 

values listed in Table 4.10 are based on values applied in previously benchmark calculations for Cooper 

fuel assembly decay heat3

 and isotopic validation studies,35 which were derived using typical BWR plant 

data. The number of burnable poison rods in the GE 7 × 7 assemblies was also not provided. However, 

two of the Cooper assemblies, designated CZ346 and CZ348, were selected for destructive radiochemical 

analysis at the Materials Characterization Center at Pacific Northwest Laboratory, and the samples 

designated as Approved Testing Material 105 (ATM-105).36 The report on ATM-105 gives the 

enrichment zoning and indicates that the assembly contained three fuel rods of 3 wt % and two fuel rods 

of 4 wt % gadolinium oxide (Gd2O3) poison, for an average of 3.4 wt %. The location of the poison rods 

in the CZ346 and CZ348 assemblies is illustrated in the assembly configuration shown in Figure 4.4. 

The same poison rod configuration and initial poison loading was assumed for all of the GE 7 × 7 

assemblies evaluated in this study. 

The Cooper assemblies had an initial assembly average enrichment of 2.50 wt % 235U, with the exception 

of assembly CZ102, which had an enrichment of 1.09 wt % 235U. The single Dresden assembly had an 

enrichment of 2.13 wt %. All Monticello assemblies were enriched to 2.25 wt % 235U. The assembly 

burnup by cycle and reactor operating data are listed in Tables 4.11–4.13. 

The activation of nonfuel structural components of the assembly, listed in (Table 4.13), was included in 

all calculations . Structural material mass was obtained from NUREG/CR-5625 (Ref. 3). These values 

were derived from data in Ref. 37 and weighted to reflect the relative flux levels in each component. 

The largest contribution of activation products to the decay heat comes from 60Co, produced by the 

activation of cobalt as an impurity. The initial cobalt levels were estimated assuming 800 ppm cobalt in 

stainless steel type 304, and 10 ppm in Zircaloy-2. The activation of non-fuel structural components of 

the assembly, listed in (Table 4.14), was included in all calculations. 
42 

Table 4.10 Cooper, Dresden, and Monticello and fuel assembly design information 

Parameter Data 

Assembly and reactor data 

Lattice geometry GE 7 × 7 

Designer General Electric 

Reactor Cooper Dresden 2 Monticello

Versions used GE-3B GE-2 GE-2B 

EIA assembly code G4607G3B G4607G2 G2307G2B 

Active fuel rod length (cm) 370.8 365.8 365.8 

Uranium mass per assembly (kg)a 190 193 193 

Assembly pitch (cm) 15.24 

Rod pitch (cm) 1.875 

Number of fuel rods 49 

Inlet temperature (K) 546 

Average moderator temperature (K) 558 

Average moderator density (g/cm3

) 0.4343 

Number of Gd poison rods 5 

Channel water density (g/cm3

) 0.7627 

Channel water temperature (K) 552 

Flow tube material Zircaloy-4 

Flow tube inside flat-to-flat (cm) 13.406 

Flow tube thickness (cm) 0.2032 

Flow tube temperature (K) 558 

 

 Fuel rod data 

Fuel material type UO2

Fuel density (% TD) 94.0 

Fuel stack density (g/cm3

) 9.73 

Gadolinium rods, wt % poison 3.4 (average) 

Fuel pellet diameter (cm) 1.212 1.237 

Fuel rod outside diameter (cm) 1.430 

Effective fuel temperature (K) 811 

Clad material Zircaloy-2 

Clad thickness (cm) 0.0940 0.0813 

Average clad temperature (K) 620 

 a Represents nominal values. 
43 

Figure 4.4 Layout for the GE 7 × 7 fuel assembly 
44 

Figure 4.5 GE 7 × 7 fuel assembly dimensions 
45 

Table 4.11 Cooper Station reactor and assembly irradiation data 

Operational data Cycle 1 Cycle 2 Cycle 3 Cycle 4 Cycle 5 Cycle 6 Cycle 7 

Startup date 7/3/1974 11/15/1976 10/18/1977 5/5/1978 5/10/1979 6/7/1980 6/7/1981 

Shutdown date 9/17/1976 9/17/1977 3/31/1978 4/17/1979 3/1/1980 4/20/1981 5/21/1982 

Operating days 807 306 164 347 296 317 348 

Downtime (days) 59 31 35 23 98 48 -- 

Assembly ID Assembly burnup (MWd/MTU) 

CZ102 9,394 11,667 

CZ147 6,773 10,381 13,490 21,231 23,421 26,709 

CZ148 5,887 10,126 13,220 20,883 23,049 26,310 

CZ182 10,710 18,421 20,412 24,464 26,824 

CZ195 5,999 10,319 12,622 21,395 23,523 26,392 

CZ205 10,298 17,712 20,699 22,563 25,344 

CZ209 10,651 18,320 21,430 23,726 25,383 

CZ211 6,770 10,339 13,462 21,197 23,387 26,668 

CZ222 10,728 18,452 21,519 23,759 26,692 

CZ225 10,672 18,355 21,455 23,489 25,796 

CZ239 10,500 18,059 21,175 24,742 27,246 

CZ246 10,964 18,858 22,052 24,065 27,363 

CZ259 6,026 10,365 12,920 21,431 23,596 26,466 

CZ264 5,938 10,213 12,782 21,404 23,601 26,496 

CZ277 6,012 10,341 12,885 21,427 23,599 26,478 

CZ286 11,071 19,042 22,218 24,916 27,141 

CZ296 5,964 10,258 13,323 20,971 22,547 26,388 

CZ302 11,637 20,016 22,174 26,594 

CZ308 11,826 20,340 21,780 23,983 25,815 

CZ311 11,236 19,325 22,322 25,314 27,392 

CZ315 11,415 19,634 22,565 24,526 26,881 

CZ318 11,387 19,585 22,537 24,902 26,568 

CZ331 12,875 18,370 21,332 

CZ337 10,861 18,681 21,594 23,806 26,720 

CZ342 11,396 19,600 22,788 24,680 27,066 

CZ346 11,651 20,039 21,920 25,203 28,048 

CZ348 11,063 19,028 22,169 24,773 27,481 

CZ351 10,741 18,475 21,287 23,859 25,753 

CZ355 11,654 20,046 21,468 23,620 25,419 

CZ357 11,136 19,153 22,060 25,023 27,140 

CZ369 11,162 19,197 21,678 23,660 26,576 

CZ370 11,511 19,799 21,946 26,342 

CZ372 11,051 19,008 21,708 25,848 

CZ379 11,142 19,165 21,907 24,286 25,925 

CZ398 11,221 19,300 22,212 24,499 27,478 

CZ415 11,058 19,020 21,726 25,863 

CZ416 11,572 19,904 22,906 25,051 27,461 

CZ429 10,878 18,711 21,610 24,842 27,641 

CZ430 10,774 18,531 21,626 24,176 26,825 

CZ433 11,350 19,522 19,522 21,639 25,977 

CZ460 10,932 18,804 21,938 24,570 26,512 

CZ466 11,012 18,940 22,148 24,290 26,077 

CZ468 11,130 19,143 21,603 24,823 26,757 

CZ472 10,806 18,587 21,272 23,173 25,957 

CZ473 11,305 19,444 22,448 24,840 26,519 

CZ498 10,853 18,666 21,362 23,584 26,482 

CZ508 12,368 21,273 21,921 24,014 26,357 

CZ515 11,003 18,924 21,637 25,737 

CZ526 10,938 18,814 21,548 24,788 27,596 

CZ528 10,996 18,913 21,604 25,715 

CZ531 10,906 18,758 21,471 24,036 26,699 

CZ536 10,996 18,913 21,604 25,715 

CZ542 11,109 19,108 22,062 24,717 26,691 

CZ545 11,311 19,454 22,675 24,856 26,668 
46 

Table 4.12 Dresden 2 reactor and assembly irradiation data 

Operational data Cycle 1 Cycle 1A Cycle 2 

Startup date 5/1/1970 8/5/1970 5/26/1971 

Shutdown date 6/6/1970 2/27/1971 2/20/1972 

Operating days 36 206 270 

Downtime (days) 60 88 -- 

Assembly ID Assembly burnup (MWd/MTU) 

DN212 299 2,112 5,279 

Table 4.13 Monticello reactor and assembly irradiation data 

Operational data Cycle 1 Cycle 2 Cycle 3 Cycle 4 

Startup date 2/19/1971 5/19/1973 5/19/1974 2/8/1975 

Shutdown date 3/3/1973 3/15/1974 1/10/1975 9/12/1975 

Operating days 743 300 236 216 

Downtime (days) 77 65 29 -- 

Assembly ID Assembly burnup (MWd/MTU) 

MT116 8,559 13,288 14,721 18,039 

MT123 8,353 13,027 

MT133 8,893 13,870 17,460 20,994 

MT190 4,998 7,893 11,349 15,143 

MT228 3,522 5,389 8,724 12,123 

MT264 3,685 5,563 7,301 9,160 
47 

Table 4.14 Activated structural elements in Cooper, Dresden, 

and Monticello assemblies 

Element grams per kg Ua

 kilograms per assemblyb

Oxygen 134 

Chromium 2.4 0.45 

Manganese 0.15 0.029 

Iron 6.6 1.2 

Cobalt 0.024 0.0046 

Nickel 2.4 0.45 

Zirconium 516.0 98.2 

Niobium 0 0 

Tin 8.7 1.6 

a

 Values derived from NUREG/CR-5625 (Ref. 3). 

b

 Assumed nominal mass of 190 kg U per assembly. 
48 

4.1.4.1 Cooper Station 

The Cooper Station operating history along with assembly data and burnup by cycle are given in 

Table 4.11. The operating data for all Cooper assemblies presented in the reports PNL-5777 (Ref. 38) 

and EPRI NP-4619 (Ref. 30) were normalized to yield with discharge data listed in RW-859 (Ref. 32). 

A review of the data confirmed that the burnup values were consistent between the different sources with 

the exception of assembly CZ536. PNL-5777 reports that the assembly was discharged after Cycle 7 with 

a burnup of 25,715 MWd/MTU, whereas RW-859 lists the discharge at Cycle 6 with a final burnup of 

26,589 MWd/MTU. This discrepancy raises concerns over whether the assembly was identified 

correctly, and moreover, the accuracy of the decay time. Based on the calculated results for assembly 

CZ536 the Cycle 7 discharge date appears to be correct (e.g., a Cycle 6 discharge date would give a 

considerably longer decay time and produce an uncharacteristic underprediction of decay heat relative to 

the other measured assemblies). For this study the assembly operating history data from Table 4.1 of 

report PNL-5777 were applied. The initial uranium mass of each assembly was obtained from RW-859. 

4.1.4.2 Dresden 2 

The reactor operating information and irradiation data for Dresden 2 assembly DN212 are listed in 

Table 4.12. The Nuclear Fuel Data from RW-859 (Ref. 32) indicates that Dresden 2 assembly DN212 

was discharged after Cycle 2. However, report EPRI NP-4619 (Ref. 30) Table 2-3 suggests the assembly 

was irradiated only in Cycle 1. Further analysis of the operating history for DN212 in Table A-3 of EPRI 

NP-4619 confirms that the assembly was likely irradiated for 2 cycles. The plant operating data in RW859 further indicates that Cycle 1 was divided in 2 parts, identified as Cycle 1 and Cycle 1A. This is also 

confirmed from the detailed operating history; however, the dates of operation and shutdown are 

inconsistent between the two references. RW-859 indicates that Cycle 1 started on 5/29/1970 and 

continued for 7 d before a shut down of 67 d prior to the start of Cycle 1A. EPRI NP-4619 gives a 

starting date of 5/01/1970 with operation for 36 d followed by a 60-d shutdown. The data for Cycle 1A 

and Cycle 2 are more consistent. For this study the operating history information for assembly DN212 

was modeled using Cycles 1, 1A, and 2 using the detailed reactor power data provided in 

Table A-3 of report EPRI NP-4619. 

4.1.4.3 Monticello 

The reactor operating information and irradiation data for the Monticello assemblies are listed in 

Table 4.13. The Monticello assembly information was obtained primarily from report PNL-5799 

(Ref. 30). The experience with most of the assemblies evaluated in this study is that the total burnup 

obtained from the cycle burnup values reported by the utilities is consistent with the discharge burnups 

reported in RW-859 (note that RW-859 documents only the final burnup, not the accumulated cycle 

burnup). However, for the Monticello assemblies this was not the case. The difference is attributed by 

McKinnon et al.30 to different accounting methods used by the utility. The difference between the burnup 

estimates obtained by the different methods was nearly 10% in some cases. On average the burnup 

reported in RW-859 was about 2% lower than the value based on cycle-by-cycle data, with an average 

relative standard deviation of ±5%. For consistency with the other evaluations in this study, all 

Monticello assembly calculations were performed by normalizing the cycle burnup data to the final 

discharge burnup reported in RW-859. 
49 

4.2 Assembly Designs—Sweden Measurements

The decay heat measurements performed at the Swedish CLAB fuel storage facility that are evaluated in 

this report include 34 assemblies from 8 BWR reactor units and 30 assemblies from 2 PWR reactor units 

operated in Sweden. The assemblies are grouped based on the assembly design. There are four BWR 

designs (8 × 8, 9 × 9, SVEA-64 and SVEA-100) and two PWR designs (15 × 15 and 17 × 17). Each 

group of assemblies can be further subdivided based on the details of the assembly configuration such as 

fuel pellet or rod diameter, number of water rods, number of burnable absorber rods, etc. Detailed 

information on the assembly design, irradiation history, and axial burnup and void distributions can be 

found in Ref. 5. 

At the time the present calculations were performed, detailed assembly enrichment zoning patterns and 

burnable poison rod configurations for each assembly, published in Ref. 5, were not available. Therefore, 

typical patterns based on past experience were applied. The assembly and fuel data for the assemblies 

analyzed in this study are summarized separately in Ref. 39. This report includes much of the design and 

operating data from Ref. 5 and also includes information on the assembly models developed for the decay 

heat calculations and derives other required data not provided in the SKB report. For completeness, the 

data relevant to the assembly models developed for this study are presented here. 

4.2.1 BWR 8 × 8 Assemblies 

The BWR 8 × 8 fuel assembly design data are given in Table 4.15. Assemblies that were rebuilt are not 

included in Table 4.15 because insufficient information was available to develop a reliable computational 

model. The specific density of the uranium oxide reported in the table for each assembly type is the 

average value over all assemblies of that type. The specific fuel density for each individual assembly can 

be found in Ref. 5. 

There were four types of 8 × 8 assembly designs studied. Two of the groups (Type 1 and 2) contain 

assemblies with a single water rod (designated as assembly type 8 × 8−1). All groups except Type 4 used 

corner rods having dimensions slightly smaller than the rods in the rest of the assembly. Another 

difference between assembly types is the number of burnable absorber rods and the content of Gd2O3 in 

these rods. Several assumptions on material and dimensions had to be made when information was not 

available. The material for the fuel clad and the water rods was assumed to be Zircaloy-2. The outer 

diameter and thickness of the water rods for some of the assembly types were assumed to be the same as 

for the fuel rod cladding in the assemblies of the same type. Type 3 and Type 4 assemblies also contained 

a single zirconium spacer rod. These spacer rods have the same outside diameter as the water rods but 

have a solid Zircaloy interior, i.e., they do not contain water. 

The BWR 8 × 8 assemblies were irradiated in seven different reactors: Ringhals 1, Oskarshamn 2, 

Oskarshamn 3, Forsmark 1, Forsmark 2, Barsebäck 1, and Barsebäck 2, and cover a burnup range from 

14 to 41 GWd/MTU. The burnup and irradiation history data for each assembly and each reactor type are 

presented in Tables 4.16–4.22. The average coolant density, the number of burnable absorber rods, and 

the weight percent of Gd2O3 in these rods, as well as the average fuel enrichment and the initial mass of 

uranium, are shown in Table 4.23. The coolant density was calculated as an assembly-averaged value 

based on the axial void distribution for each assembly provided in Ref. 5. For assemblies from the reactor 

Ringhals-1, axial void information was available only for the assemblies identified as 1177 and 1186. 

The average of the coolant density for these two assemblies was used for the other 8 × 8 assemblies from 

the same reactor. Because specific assembly layout information data was not available at the time these 

calculations were performed, the location of the water rods in the 8 × 8 assemblies was based on typical 

data from other sources.40 Assumptions were also made regarding the location of the burnable absorber 
50 

rods in the assembly; more details on these assumptions are discussed when presenting the actual 

computational models for each assembly type in Sect. 5. A typical 8 × 8 assembly layout used in the 

computational model is provided in Figure 4.6. 
51 

Table 4.15 Fuel assembly design data for BWR 8 × 8 assemblies 

Data Parameter 

Type 1 Type 2 Type 3 Type 4 

Assembly data 

Lattice geometry 8 × 8−1 8 × 8−1 8 × 8−1 8 × 8−2 

Fuel rod active length (cm) 365.0 368.0 371.2 368.0 

Assembly pitch (cm) 15.33 15.33 a 15.30 15.38 

Moderator b

 temperature (K) 552 552 552 552 

Moderator b density (g/cm3

) 0.75 0.75 0.75 0.75 

Coolant temperature (K) 552 552 552 552 

Number of normal fuel rods 51 51 51 62 

Number of corner fuel rods 12 12 12 N/A 

Number of water rods 1 1 0 1 

Number of Zr spacer rods 0 0 1 1 

Rod pitch (cm) 1.63 1.63 1.63 1.625 

Rod pitch for corner rods (cm) 1.58 1.58 1.58 N/A 

Rod pitch for normal-to-corner rods (cm) 1.605 1.605 1.605 N/A 

Assembly channel outer dimension (cm) 13.9 13.9 13.86 13.9 

Assembly channel wall thickness (cm) 0.2 0.23 0.23 0.23 

Assembly channel material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 

Spacer material c Inconel Inconel Inconel Inconel 

Spacer total mass (g) 810 810 810 1938 

Number of absorber rods d 3 0/3/4 0/4/5/6 4 

Gd2O3 wt % in absorber rods d 2 2/3.95/5.5 2/2.55/3.2 2.5 

 

Fuel rod data 

Fuel material type UO2 UO2 UO2 UO2

Effective fuel density (g/cm3) 10.168 10.316 10.621e/10.429f 10.171 

Effective fuel temperature (K) 900 900 900 900 

Fuel pellet diameter, normal rods (cm) 1.058 1.044 1.044 1.044 

Fuel pellet diameter, corner rods (cm) 1.008 0.994 0.994 N/A 

Fuel rod outside diameter, normal rods (cm) 1.225 1.225 1.225 1.23 

Fuel rod outside diameter, corner rods (cm) 1.175 1.175 1.175 N/A 

Clad material g Zircaloy-2 Zircaloy-2 Zircaloy-2 Zircaloy-2 

Clad thickness for normal rods (cm) 0.074 0.08 0.08 0.082 

Clad thickness for corner rods (cm) 0.074 0.08 0.08 N/A 

Average clad temperature (K) 573 573 573 573 

 

Water rod data 

Water rod material g Zircaloy-2h Zircaloy-2h Zircaloy-2 Zircaloy-2 

Water rod outer diameter (cm) 1.225h 1.225h 1.225 1.5 

Water rod thickness (cm) 0.074h 0.074h 0.074h 0.08 

a

 Some assemblies of this type have different values; assembly pitch for assemblies with identifiers 3838, 5535 and 

12078 have a pitch value of 15.38 cm, 15.40 cm, and 15.50 cm, respectively. b

 This is the moderator in the assembly channel. 

c

 Inconel X-750 is assumed. 

d

 Number varies with assembly; exact values are shown in Table 4.23. 

e

 For assemblies with no absorber rods. 

f

 For assemblies with absorber rods. 

g

 Zircaloy-2 is assumed. 

h

 Assumed. 
52 

52 

Table 4.16 Ringhals-1 reactor and 8 × 8 assembly irradiation history data 

Operational data type Cycle 2 Cycle 3 Cycle 4 Cycle 5 Cycle 6 Cycle 7 Cycle 8 Cycle 9 Cycle 10 Cycle 11 Cycl

Startup date 10/13/1978 9/30/1979 10/8/1980 11/6/1981 9/9/1982 9/30/1983 8/28/1984 9/2/1985 10/9/1986 9/24/1987 9/11Shutdown date 7/12/1979 7/3/1980 7/30/1981 7/22/1982 6/18/1983 7/13/1984 8/2/1985 8/15/1986 8/21/1987 8/6/1988 9/15Operating days 272 277 295 258 282 287 339 347 316 317 Downtime days 80 97 99 49 104 46 31 55 34 36 

Assembly ID Design Assembly Burnup (MWd/MTU) 

1177 8 × 8 4,486 9,369 16,334 20,430 25,484 30,320 36,242 1186 8 × 8 4,554 9,614 15,459 19,967 24,592 28,020 30,498 6423 8 × 8 5,401 12,064 19,440 26,198 31,635 35,109 6432 8 × 8 5,502 10,444 16,181 23,266 30,079 34,746 36,861 6454 8 × 8 7,120 13,000 18,936 24,831 30,918 37,236 8327 8 × 8 3,212 9,482 15,956 21,578 28,065 38331 8 × 8 6,974 15,132 22,493 26,648 33,319 38332 8 × 8 5,755 11,468 16,558 23,407 29,963 32,651 34,977 8338 8 × 8 5,660 11,130 16,282 23,202 29,963 32,634 34,830 

Table 4.17 Oskarshamn-2 reactor and 8 × 8 assembly irradiation data 

Operational data type Cycle 1 Cycle 2 Cycle 3 Cycle 4 Cycle 5 Cycle 6 Cycle 7 Cycle 8 Cycle 9 Startup date 10/2/1974 7/2/1976 8/7/1977 9/3/1978 8/30/1979 8/20/1980 8/30/1981 8/23/1982 9/19/1983 Shutdown date 4/16/1976 5/13/1977 6/24/1978 7/20/1979 7/12/1980 7/15/1981 7/23/1982 8/19/1983 7/1/1984 Operating days 562 315 321 320 317 329 327 361 286 Downtime days 77 86 71 41 39 46 31 31 54 Assembly ID Type Assembly Burnup (MWd/MTU) 

1377 3 8,982 14,546 1389 3 4,315 7,300 11,829 14,539 16,933 19,481 1546 3 6,961 13,840 21,032 24,470 1696 3 4,492 7,904 13,258 16,319 18,502 20,870 1704 3 4,587 8,031 13,384 16,638 19,437 2995 3 6,143 13,359 19,705 23,577 29,978 6350 3 7,113 14,909 24,019 
53 

53 

Table 4.18 Forsmark-1 reactor and 8 × 8 assembly irradiation data 

Operational data type Cycle 1A Cycle 1B Cycle 2 Cycle 3 Cycle 4 Cycle 5 Cycle 6 Cycle 7 Cycle 8 Startup date 4/23/1980 8/16/1981 8/25/1982 8/5/1983 8/6/1984 7/8/1985 7/22/1986 8/19/1987 7/10/1988 8Shutdown date 6/25/1981 6/19/1982 7/15/1983 7/13/1984 5/31/1985 7/4/1986 7/31/1987 6/10/1988 7/14/1989 8

Operating days 428 307 324 343 298 361 374 296 369 Downtime days 52 67 21 24 38 18 19 30 35 Assembly ID Type Assembly Burnup (MWd/MTU) 

3838 2 6,160 13,092 15,990 19,053 

KU0100 4 10,643 18,706 26,630 34,193 

Table 4.19 Forsmark-2 reactor and 8 × 8 assembly irradiation data

Operational data type Cycle 1A Cycle 1B Cycle 2 Cycle 3 Cycle 4 Cycle 5 Cycle 6 Startup date 12/1/1980 8/18/1982 6/11/1983 7/13/1984 7/26/1985 8/9/1986 6/20/1987 Shutdown date 5/1/1982 5/20/1983 5/25/1984 7/5/1985 7/25/1986 5/31/1987 7/15/1988 Operating days 516 275 349 357 364 295 391 Downtime days 109 22 49 21 15 20 --- Assembly ID Type Assembly Burnup (MWd/MTU) 

5535 2 8,850 10,758 12,862 14,982 16,665 18,115 19,944 

Table 4.20 Barsebäck-1 reactor and 8 × 8 assembly irradiation data

Operational data type Cycle 8 Cycle 9 Cycle 10 Cycle 11 Cycle 12 Startup date 8/6/1983 7/26/1984 9/2/1985 8/18/1986 7/29/1987 Shutdown date 6/28/1984 8/7/1985 7/17/1986 7/1/1987 9/17/1988 Operating days 327 377 318 317 416 Downtime days 28 26 32 28 --- Assembly ID Type Assembly Burnup (MWd/MTU) 

9329 3 8,219 17,727 25,034 31,958 41,127 

10288 3 9,726 17,440 25,587 35,218 
54 

Table 4.21 Barsebäck-2 reactor and 8 × 8 assembly irradiation data 

Operational 

data type Cycle 9 Cycle 10 Cycle 11 Cycle 12 Cycle 13 

Startup date 9/19/1987 8/7/1988 9/28/1989 8/18/1990 9/22/1991

Shutdown date 7/6/1988 9/8/1989 7/11/1990 9/6/1991 7/2/1992

Operating days 291 397 286 384 284

Downtime days 32 20 38 16 --

Assembly ID Type Assembly Burnup (MWd/MTU) 

14076 3 8,398 18,616 24,130 34,126 40,010

Table 4.22 Oskarshamn-3 reactor and 8 × 8 assembly irradiation history 

Operational 

data type Cycle 1 Cycle 2 Cycle 3 

 Startup date 3/18/1985 7/26/1986 7/26/1987 

 Shutdown date 7/4/1986 7/3/1987 7/8/1988 

 Operating days 473 342 348 

 Downtime days 22 23 37 

Assembly ID Design Assembly Burnup (MWd/MTU) 

12078 8 × 8 8,690 16,948 25,160 
55 

Table 4.23 Other data for BWR 8 × 8 assemblies 

Assembly ID Type Reactor Enrichment 

(wt % 235U) 

Initial U 

(kg) 

Coolant 

density 

(g/cm3

) a

No. 

BA

rods b

Gd2O3

in BA 

(wt %) 

1177 1 Ringhals 1 2.642 180.58 0.4325 3 2 

1186 1 Ringhals 1 2.640 180.52 0.4596 3 2 

6423 2 Ringhals 1 2.900 177.70 0.4445 4 2 

6432 2 Ringhals 1 2.894 177.52 0.4445 4 2 

6454 2 Ringhals 1 2.898 177.68 0.4445 4 2 

8327 2 Ringhals 1 2.904 177.54 0.4445 4 2 

8331 2 Ringhals 1 2.910 177.69 0.4445 4 2 

8332 2 Ringhals 1 2.895 177.52 0.4445 4 2 

8338 2 Ringhals 1 2.911 177.60 0.4445 4 2 

1377 3 Oskarshamn 2 2.201 183.58 0.3790 0 N/A 

1389 3 Oskarshamn 2 2.201 183.65 0.5053 0 N/A 

1546 3 Oskarshamn 2 2.201 183.97 0.4240 0 N/A 

1696 3 Oskarshamn 2 2.201 184.25 0.4894 0 N/A 

1704 3 Oskarshamn 2 2.201 184.02 0.4622 0 N/A 

2995 3 Oskarshamn 2 2.699 179.38 0.4135 4 2 

6350 3 Oskarshamn 2 2.875 179.00 0.3967 6 3.2 

12078 2 Oskarshamn 3 2.578 177.36 0.3946 3 5.5 

9329 3 Barsebäck 1 2.922 178.77 0.4199 5 2.55 

10288 3 Barsebäck 1 2.953 179.16 0.4029 5 2.55 

14076 3 Barsebäck 2 3.15 179.57 0.3928 4 2 

3838 2 Forsmark 1 2.090 177.90 0.5073 3 3.95 

KU0100 4 Forsmark 1 2.970 174.92 0.4001 4 2.5 

5535 2 Forsmark 2 2.095 177.69 0.5167 0 N/A 

a Axial volumetric average. 

b BA – burnable absorber rod. 
56 

Figure 4.6 Layout of a BWR 8 × 8-1 assembly 
57 

4.2.2 BWR 9 × 9 Assemblies 

Decay heat measurements were performed for three BWR 9 × 9 assemblies from the Forsmark-1 reactor. 

The assemblies have a total burnup in the range 35–38 GWd/MTU. The fuel assembly design data are 

presented in Table 4.24. Information on the Forsmark reactor operating history, assembly irradiation 

data, initial uranium mass, initial enrichment, and coolant density are shown in Table 4.25. Each 

assembly has 76 fuel rods, 4 water rods, and 1 solid Zr spacer rod. Six of the fuel rods contain a burnable 

absorber. The positions of the water rods and burnable absorber rods in the assembly were not available 

at the time, and were therefore based on design information for a similar KWU (Siemens) 9 × 9−5 

assembly design used in the Barsebäck reactor. The assembly layout is shown in Figure 4.7. 

Table 4.24 Fuel assembly design data for BWR 9 × 9 assemblies 

Parameter Data 

Assembly and reactor data 

Lattice geometry 9 × 9 

Fuel rod active length (cm) 368.0 

Assembly pitch (cm) 15.38 

Channel moderator temperature (K) 552 

Channel moderator density (g/cm3

) 0.75 

Coolant temperature (K) 552 

Number of fuel rods 76 

Number of water rods 4 

Number of Zr spacer rods 1 

Rod pitch (cm) 1.445 

Assembly channel outer dimension (cm) 13.9 

Assembly channel wall thickness (cm) 0.23 

Assembly channel material Zircaloy-4 

Spacer material Zircaloy-4 

Spacer total mass (g) 1938 

Number of absorber rods 6 

Gd2O3 wt % in absorber rods 2.5 

Fuel rod data 

Fuel material type UO2

Effective fuel density (g/cm3

) 10.144 

Effective fuel temperature (K) 900 

Fuel pellet diameter (cm) 0.95 

Fuel rod outside diameter (cm) 1.1 

Clad material a Zircaloy-2 

Clad thickness (cm) 0.0665 

Average clad temperature (K) 573 

Water rod data 

Water rod material a Zircaloy-2 

Water rod outer diameter (cm) 1.445 

Water rod thickness (cm) 0.08 

a

 Not specified; Zircaloy-2 is assumed. 
58 

Table 4.25 Forsmark 1 reactor and 9 × 9 assembly irradiation data 

Operational data Cycle 5 Cycle 6 Cycle 7 Cycle 8 Cycle 9

Startup date 7/8/1985 7/22/1986 8/19/1987 7/10/1988 8/18/1989

Shutdown date 7/4/1986 7/31/1987 6/10/1988 7/14/1989 8/17/1990

Operating days 361 374 296 369 364

Downtime days 18 19 30 35 

 

Assembly data 

ID Enrichment 

(wt % 235U) 

Initial U 

(kg) 

Coolant 

density 

(g/cm3

) 

Assembly Burnup (MWd/MTU) 

KU0269 2.938 177.02 0.3866 10,158 18,787 27,153 35,113 

KU0278 2.938 177.13 0.4235 10,414 17,685 24,764 29,413 35,323

KU0282 2.938 177.10 0.4247 10,122 18,800 25,844 32,112 37,896
59 

Figure 4.7 Layout of a BWR 9 × 9−5 assembly
60 

4.2.3 BWR SVEA-64 Assemblies 

Decay heat measurements were performed for four SVEA-64 assemblies: three irradiated in the 

Forsmark-2 reactor and one irradiated in the Oskarshamn-2 reactor. Each assembly contains four 4 × 4 

subassemblies separated by a large water region (water cross). There are 63 fuel rods and 1 water rod in 

each assembly. Some of the fuel rods (4 to 6, depending on the assembly type) contain Gd2O3 burnable 

poison. Information on the assembly design is presented in Table 4.26, and data on irradiation history, 

coolant density, fuel enrichment and initial uranium mass is given in Tables 4.27 and 4.28. The assembly 

layout is shown in Figure 4.8 for an assembly with four burnable absorber rods. 

Table 4.26 Fuel assembly design data for SVEA-64 assemblies 

Parameter Data 

Assembly and reactor data Type 1 

Lattice geometry 8 × 8 

Assembly pitch (cm) 15.40 

Channel moderator temperature (K) 552 

Channel moderator density (g/cm3

) 0.72 

Coolant temperature (K) 552 

Number of fuel rods 63 

Number of water rods 1 

Rod pitch (cm) 1.58 

Assembly channel outer dimension (cm) 14.0 

Assembly channel wall thickness (cm) 0.11 

Assembly channel material Zircaloy-4 

Water cross material Zircaloy-4 

Water cross wall thickness (cm) 0.08 

Sub-assembly inner measure a

 (cm) 6.59 

Spacers material b Inconel 

Spacers total mass (g) 840/744 c

Number of absorber rods 4/5/6 d

Gd2O3 wt % in absorber rods 2.55/2.55/3.15 

Fuel rod data 

Fuel material type UO2

Effective fuel density (g/cm3

) 10.367 

Effective fuel temperature (K) 900 

Fuel pellet diameter (cm) 1.044 

Fuel rod outside diameter (cm) 1.225 

Clad material e Zircaloy-2 

Clad thickness (cm) 0.08 

Average clad temperature (K) 573 

Water rod data 

Water rod material Zircaloy-2 

Water rod outer diameter (cm) 1.225 

Water rod thickness (cm) 0.08 a

 Wall to wall distance across the subassembly box. b

 Not specified; Inconel is assumed. c Spacers total mass is 840 g for assemblies 11494, 11495, and 13775; 

744 g for assembly 12684. 

d

 4 rods for assemblies 11494 and 11495; 5 rods for assembly 13775; 6 

rods for assembly 12684. 

e

 Zircaloy-2 is assumed.
61 

61 

Table 4.27 Forsmark-2 reactor and SVEA-64 assembly irradiation data 

Operational data Cycle 3 Cycle 4 Cycle 5 Cycle 6 Cycle 7 CyStartup date 7/13/1984 7/26/1985 8/9/1986 6/20/1987 8/4/1988 9/4/Shutdown date 7/5/1985 7/25/1986 5/31/1987 7/15/1988 8/18/1989 5/18/Operating days 357 364 295 391 379 Downtime days 21 15 20 20 17 Assembly data 

ID Type Enrichment (wt % 235U) Initial U (kg) Coolant density (g/cm3) Assembly Burnup (MWd/MTU) 

11494 1 2.920 181.09 0.4348 9,098 16,676 24,136 32,431 11495 1 2.910 181.07 0.4106 9,098 16,676 24,136 32,431 13775 1 2.850 181.34 0.4449 8,801 18,300 23,975 27,4

Table 4.28 Oskarshamn-2 reactor and SVEA-64 assembly irradiation data 

Operational data Cycle 11 Cycle 12 Cycle 13 Cycle 14 Cycle 15 Startup date 7/13/1985 9/27/1986 9/4/1987 9/10/1988 9/1/1989 Shutdown date 8/16/1986 7/31/1987 8/20/1988 8/5/1989 8/10/1990 Operating days 399 307 351 329 343 Downtime days 42 35 21 27 62 Assembly data 

ID Type Enrichment (wt % 235U) Initial U (kg) Coolant density (g/cm3) Assembly Burnup (MWd/MTU) 

12684 1 2.902 182.32 0.3762 11,134 20,032 28,184 35,495 43,084 
62 

Figure 4.8 Layout of a BWR SVEA-64 assembly 
63 

4.2.4 BWR SVEA-100 Assemblies 

Decay heat measurements were performed for four SVEA-100 assemblies: two of these assemblies were 

from the Oskarshamn-3 reactor and two from the Forsmark-3 reactor. The assemblies had a burnup range 

of about 31–40 GWd/MTU. Each SVEA-100 assembly contains four 5 × 5 sub-assemblies divided by a 

large water cross. There are 100 fuel rods in an assembly, some of these (5 or 8, depending of the 

assembly type) containing burnable poison. The assemblies contained no water rods or spacer rods. Data 

on the assembly design is presented in Table 4.29. Information on the reactor operating history and 

assembly irradiation, the average coolant density, fuel enrichment and initial uranium mass are given in 

Tables 4.29 and 4.30. The assembly layout is shown in Figure 4.9 for an assembly with six burnable 

absorber rods. 

Table 4.29 Fuel assembly design data for SVEA-100 assemblies 

Parameter Data 

Assembly and reactor data Type 1 

Lattice geometry 10 × 10 

Assembly pitch (cm) 15.50 

Channel moderator temperature (K) 552 

Channel moderator density (g/cm3

) 0.75 

Coolant temperature (K) 552 

Number of fuel rods 100 

Rod pitch (cm) 1.24 

Assembly channel outer dimension (cm) 13.96 

Assembly channel wall thickness (cm) 0.11 

Assembly channel material Zircaloy-4 

Water cross material Zircaloy-4 

Water cross wall thickness (cm) 0.11 

Sub-assembly inner dimension a

 (cm) 6.56 

Spacer material Inconel 

Total spacer mass (g) 576/840 c

Number of burnable absorber rods 5/8 d

Gd2O3 wt % in absorber rods 2.55/4.4 

 

Fuel rod data 

Fuel material type UO2

Effective fuel density (g/cm3

) 10.392 

Effective fuel temperature (K) 900 

Fuel pellet diameter (cm) 0.819 

Fuel rod outside diameter (cm) 0.962 

Clad material b Zircaloy-2 

Clad thickness (cm) 0.063 

Average clad temperature (K) 573 

a Flat-to-flat distance across the subassembly box. b

 Zircaloy-2 material is assumed. 

c

 Total spacer mass is 576 g in assemblies 13620 and 13630; 840 g in 

assemblies 13847 and 13848. 

d

 There are 5 absorber rods in assemblies 13847 and 13848; 8 absorber 

rods in assemblies 13628 and 13630. 
64 

Table 4.30 Oskarshamn-3 reactor and SVEA-100 assembly irradiation data 

Operational data Cycle 2 Cycle 3 Cycle 4 Cycle 4B Cycle 5 

Startup date 7/26/1986 7/26/1987 8/14/1988 12/22/1988 6/20/1989 Shutdown date 7/3/1987 7/8/1988 12/10/1988 6/7/1989 6/23/1990 Operating days 342 348 118 167 368 Downtime days 23 37 12 13 39 Assembly data 

ID Type Enrichment (wt % 235U) Initial U (kg) Coolant density (g/cm3) Assembly Burnup (MWd/t) 

13628 1 2.711 180.77 0.3940 9,171 18,613 20,910 24,673 32,885 13630 1 2.711 180.78 0.3824 9,328 18,984 21,970 25,793 32,866 

Table 4.31 Forsmark-3 reactor and SVEA-100 assembly irradiation data 

Operational data Cycle 2 Cycle 3 Cycle 4 Cycle 5 Startup date 9/2/1986 8/9/1987 9/5/1988 7/13/1989 Shutdown date 7/10/1987 8/12/1988 6/9/1989 7/13/1990 Operating days 311 369 277 365 Downtime days 30 24 34 --- Assembly data 

ID Type Enrichment (wt % 235U) Initial U (kg) Coolant density (g/cm3) Assembly Burnup (MWd/t) 

13847 1 2.770 180.67 0.3816 7,857 17,217 23,779 31,275 13848 1 2.770 180.67 0.3798 7,857 17,217 23,779 31,275 
65 

Figure 4.9 Layout of a BWR SVEA-100 assembly 
66 

4.2.5 PWR 15 × 15 Assemblies 

Decay heat measurements were performed for sixteen 15 × 15 PWR assemblies that have been irradiated 

in the Ringhals 2 reactor. Each assembly has 204 fuel rods, 20 guide tubes and 1 instrument tube. 

No burnable absorber rods were present in the assembly. The assemblies are grouped into two types 

based on the dimensions of the fuel pellet, clad thickness, guide tube diameter, spacer material and total 

spacer mass. The assembly design data are presented in Table 4.32. The fuel density for each assembly 

type is an average value over all assemblies of that type. The irradiation history information for each 

assembly is presented in Table 4.33, and the fuel enrichment and initial uranium load are given in 

Table 4.34. The total burnup of the Ringhals 15 × 15 assemblies is in the range 28–51 GWd/MTU. 

The configuration of the assembly is shown in Figure 4.10. 

Table 4.32 Fuel assembly design data for PWR 15 × 15 assemblies 

Parameter Data 

Assembly and reactor data Type 1 Type 2 

Lattice geometry 15 × 15 15 × 15 

Fuel rod active length (cm) 365.8 365.8 

Assembly pitch (cm) 21.50 21.50 

Coolant density (g/cm3

) 0.72 0.72 

Coolant temperature (K) 552 552 

Average soluble boron level (ppm) 650 650 

Number of fuel rods 204 204 

Number of guide tubes 20 20 

Number of instrument tubes 1 1 

Rod pitch (cm) 1.43 1.43 

Spacers material Inconel Inconel/Zircaloy-4 

Spacers total mass (g) 788 720/160 

 

Fuel rod data 

Fuel material type UO2 UO2

Effective fuel density (g/cm3

) 10.227 10.137 

Effective fuel temperature (K) 900 900 

Fuel pellet diameter (cm) 0.929 0.911 

Fuel rod outside diameter (cm) 1.072 1.075 

Clad material Zircaloy-4 Zircaloy-4 

Clad thickness (cm) 0.0618 0.0725 

Average clad temperature (K) 573 573 
67 

Table 4.33 Ringhals-2 reactor and PWR 15 × 15 assembly irradiation data

Operational data Cycle 1 Cycle 2 Cycle 3 Cycle 4 Cycle 5 Cycle 6 Cycle 7 Cycle 8 Cycle 9 Cycle 10 Startup date 6/19/1974 7/7/1977 5/26/1978 6/25/1979 6/18/1980 6/23/1981 7/29/1982 7/28/1983 7/12/1984 6/14/1985 

Shutdown date 

4/13/1977 3/31/1978 4/3/1979 4/1/1980 4/4/1981 5/6/1982 4/28/1983 4/13/1984 4/4/1985 4/30/1986 

Operating days 

1029 267 312 281 290 317 273 260 266 320 

Downtime days 

85 56 83 78 80 84 91 90 71 63 

Assembly data 

ID Type Assembly Burnup (MWd/MTU) 

C01 1 11,247 20,650 28,219 36,688 C12 1 11,247 20,565 27,955 36,385 C20 1 11,247 20,624 28,078 35,720 

D27 2 9,510 22,399 31,666 39,676 D38 2 6,367 15,698 23,056 31,757 39,403 

E38 2 7,568 16,026 25,905 33,973 E40 2 7,705 14,954 25,609 34,339 

F14 2 5,069 15,824 25,722 34,009 F21 2 4,767 11,084 21,130 29,385 36,273 F25 2 8,307 19,056 27,372 35,352 F32 2 10,553 21,162 29,553 37,314 

G11 2 6,890 17,312 25,180 32,123 35,463 G23 2 10,268 20,303 27,921 35,633 

I09 2 6,727 15,677 24,742 I24 2 8,245 17,212 26,356 34,294 I25 2 5,207 10,198 20,001 28,999 
68 

Table 4.34 Fuel enrichment and initial uranium load for 

PWR 15 × 15 assemblies 

Assembly ID Type Enrichment 

(wt % 235U) 

Initial U 

(kg) 

C01 1 3.095 455.79 

C12 1 3.095 453.74 

C20 1 3.095 454.76 

D27 2 3.252 432.59 

D38 2 3.252 434.21 

E38 2 3.199 433.59 

E40 2 3.199 434.24 

F14 2 3.197 436.38 

F21 2 3.197 435.94 

F25 2 3.197 437.29 

F32 2 3.197 436.99 

G11 2 3.188 436.18 

G23 2 3.206 436.13 

I09 2 3.203 437.35 

I24 2 3.203 429.60 

I25 2 3.203 433.06 
69 

Figure 4.10 Layout of Ringhals PWR 15 × 15 assembly 
70 

4.2.6 PWR 17 × 17 Assemblies 

Decay heat measurements data are available for fourteen 17 × 17 PWR assemblies that have been 

irradiated in the Ringhals 3 reactor. Each assembly contains 264 fuel rods, 24 guide tubes, and 

1 instrument tube. Four of these assemblies (with identifiers 0C9, 1C5, 3C5 and 4C7) were exposed to 

burnable absorber rods located inside 12 of the 24 guide tubes. The absorber rods employ borosilicate 

glass as the absorber material in an annular rod configuration with inner void. The configuration of the 

12 absorber rods was obtained from NUREG/CR-6761 (Ref. 41). The total burnup of the assemblies is in 

the range 20–42 GWd/MTU. Assembly design data are shown in Table 4.35. The fuel density shown in 

Table 4.35 is an average value over all assemblies. Information on irradiation history data, fuel 

enrichment and initial uranium load is shown in Table 4.36. A layout for an assembly containing 

burnable absorber rods is shown in Figure 4.11. 

Table 4.35 Fuel assembly design data for PWR 17 × 17 assemblies 

Parameter Data 

Assembly and reactor data 

Lattice geometry 17 × 17 

Fuel rod active length (cm) 365.8 

Assembly pitch (cm) 21.50 

Coolant density (g/cm3

) 0.72 

Coolant temperature (K) 552 

Average soluble boron level (ppm) 650 

Number of fuel rods 264 

Number of guide tubes 24 

Number of instrument tubes 1 

Rod pitch (cm) 1.26 

Spacers material Inconel 718 

Total spacer mass (g) 788 

Fuel rod data 

Fuel material type UO2

Effective fuel density (g/cm3

) 10.27 

Effective fuel temperature (K) 900 

Fuel pellet diameter (cm) 0.8191 

Fuel rod outside diameter (cm) 0.95 

Clad material b Zircaloy-4 

Clad thickness (cm) 0.0571 

Average clad temperature (K) 573 

Guide (instrument) tube data 

Tube material Zircaloy-4 

Outer diameter (cm) 1.224 

Thickness (cm) 0.0406 

Burnable absorber rod data 

Number of rods inserted 12 

Absorber material B2O3-SiO2

Material density (g/cm3

) 2.299 

Burnable poison content (wt % B2O3) 12.5 

Outer diameter of absorber region (cm) 0.85344 

Inner diameter of absorber region (cm) 0.4826 

Outer diameter of outer clad (cm) 0.96774 

Inner diameter of outer clad (cm) 0.87376 

Outer diameter of inner clad (cm) 0.46101 

Inner diameter of inner clad (cm) 0.42799 
71 

Table 4.36 Ringhals-3 reactor and PWR 17 × 17 assembly irradiation data 

Operational data Cycle 1A Cycle 1B Cycle 2 Cycle 3 Cy

Startup date 7/29/1980 9/14/1983 7/24/1984 7/11/1985 7/18Shutdown date 6/2/1983 5/11/1984 5/25/1985 5/30/1986 6/18Operating days 1038 240 305 323 Downtime days 104 74 47 49 Assembly data 

ID BA rod 

exposure 

Enrichment 

(wt % 235U) 

Initial U 

(kg) Assembly burnup (MWd/MTU) 

2A5 No 2.100 462.03 12,228 20,107 5A3 No 2.100 461.48 11,696 19,699 

0C9 Yes 3.101 457.64 9,884 18,076 28,426 38,4421C2 No 3.101 459.05 6,249 11,268 22,777 33,3181C5 Yes 3.101 457.99 9,884 17,986 28,397 38,4842C2 No 3.101 459.49 7,783 16,128 26,060 36,5773C1 No 3.101 458.43 7,783 16,124 26,055 36,5723C5 Yes 3.101 458.87 9,884 17,997 28,340 38,3733C9 No 3.101 459.14 7,783 16,160 26,036 36,5604C4 No 3.101 459.05 6,249 11,240 22,270 33,3334C7 Yes 3.101 458.26 9,884 17,985 28,332 38,370

0E2 No 3.103 463.60 7,496 20,530 30E6 No 3.103 461.77 12,490 21E5 No 3.103 463.90 10,556 2
72 

Figure 4.11 Layout of Ringhals 3 PWR 17 × 17 assembly (with burnable absorbers) 
73 

5. COMPUTATIONAL METHODS AND MODELS 

The computational methodology used to analyze the decay heat measurements is described in this section. 

All decay heat calculations presented in this report were performed using the ORIGEN-S code. The 

methodology involved two steps: the first step was the generation of cross-section libraries for each of the 

different fuel assembly designs; the second step was the application of these libraries for the ORIGEN-S 

depletion and decay analyses to predict assembly decay heat. Cross-section libraries for the different fuel 

assembly design configurations were created using two-dimensional neutron transport methods available 

in the SCALE 5 code system. Calculations using one-dimension depletion methods were also performed 

for selected assemblies for comparison. A general description of the codes and the computational steps in 

this procedure are presented. Specific details on the models developed for each assembly type discussed 

in this study, and input examples, are also provided. 

5.1 SCALE Depletion Methods and Data

SCALE is a computational system of codes and data libraries suitable for analysis of nuclear criticality 

safety, radiation shielding, reactor lattice physics, and spent nuclear fuel characterization. SCALE is used 

internationally for nuclear fuel facility, and transportation, storage facility design, and safety analyses. 

The configuration of SCALE used in this study is the version distributed as SCALE 5.1. 

5.1.1 ORIGEN-S Code 

ORIGEN-S is the depletion module in SCALE used to calculate the isotopic composition of spent fuel, 

the neutron and gamma radiation source intensity and energy spectra, and the decay heat released from 

the decay of actinides, fission products, and activated structural components. ORIGEN-S tracks the timedependent concentrations and decay properties for 1946 individual nuclides. The decay properties of 

spent fuel are determined from the summation of the contributions from all nuclides in the problem. 

The numerical methods used in ORIGEN-S to solve the large system of coupled differential equations 

that describe the isotopic transmutation and decay processes are largely unchanged from other versions of 

the ORIGEN code.42,43 The matrix exponential method is employed to solve the nuclide transmutation 

equations as defined by a generalized matrix of transition constants. To mitigate numerical roundoff error 

the short-lived radioisotopes, as defined relative to the time step size, are removed from the transition 

matrix and are solved analytically using a generalized form of the Bateman equations that are solved by 

the Gauss-Seidel iterative method. 

The physical geometry and layout of the assembly (e.g., fuel diameter, rod pitch, water rods 

configuration) has an important effect on the flux spectrum in the fuel, and thus the average value of the 

cross sections used in depletion simulations. Reactor operating conditions (e.g., fuel temperatures, 

moderator temperatures, moderator density, soluble boron level, and control absorbers) also have a strong 

influence on the cross sections. As the fuel is irradiated the concentrations of the fissile actinides and 

fission product absorbers change, resulting also in a time dependence of the flux spectrum and effective 

cross sections. 

One of the major advantages of using SCALE for spent fuel analysis is that burnup simulations can be 

performed within control modules that automatically couple ORIGEN-S with neutronic transport codes to 

dynamically generate time-dependent cross sections for the burnup analysis. The control modules for the 

depletion analysis sequences contain automated procedures to generate time- and problem-dependent 

cross-section libraries. This allows a dramatic improvement in flexibility compared with other versions 

of the ORIGEN code, which provided the user with a limited set of precalculated libraries. For 

calculations involving fuel types or configurations not covered by available libraries, the user generally 
74 

had no access to appropriate cross-section data. The SCALE system allows the user to set up physics 

models of the fuel lattice and apply problem-dependent reactor conditions and operating history 

information to the depletion analysis. 

5.1.2 Depletion Analysis Methodology 

The SCALE depletion sequences perform automated problem-dependent burnup and decay analyses by 

automatically coupling reactor physics lattice codes to the fuel depletion code ORIGEN-S. Two 

depletion analysis sequences are available in SCALE 5.1: a 1-D depletion sequence SAS2H44 that uses 

the XSDRNPM discrete ordinates transport code,45 and TRITON depletion sequence46 that uses the 

NEWT 2-D discrete ordinates lattice code,47 with an option to perform 3-D depletion calculations using 

the KENO-V.a or KENO-VI Monte Carlo transport codes. This section provides an overview of the 1- 

and 2-D depletion analysis methods used in this study. 

Both sequences use the SCALE Material Information Processor to calculate the material number densities 

and prepare geometry data for resonance self-shielding and to create the input files for each of the crosssection processing codes. The problem-specific (resonance and temperature corrected) cross sections to 

be used with the transport codes are prepared by the BONAMI and NITAWL cross-section processing 

codes. BONAMI applies the Bondarenko method of resonance self-shielding for nuclides that have 

Bondarenko data included in the library. The SCALE libraries typically use Bondarenko data to process 

group cross sections in the unresolved resonance range. NITAWL uses the Nordheim integral method to 

perform analytical resonance self-shielding corrections for nuclides with resolved resonance parameters 

in the library. An alternate self-shielding cross section module is available in SCALE 5 that uses the 

CENTRM code (Continuous Energy Transport Module) to calculate group-average cross sections. 

CENTRM computes the continuous-energy neutron flux spectrum for a 1-D system using a combination 

of pointwise and multigroup transport cross sections. The problem-specific neutron flux is computed on a 

fine energy mesh (> 10,000 points) and can be used to generate self-shielded group cross sections from 

continuous-energy cross sections using the PMC (Produce Multigroup Cross Sections) module. 

All depletion analysis calculations performed in the present study using the NITAWL self-shielding 

module was used for resonance processing. Approximately 250 room-temperature critical benchmarks 

were used to validate the CENTRM/PMC methodology. However, there is only limited experience with 

using CENTRM/PMC for reactor operating condition and fuel depletion. The NITAWL module was 

used based on its demonstrated past performance for LWR fuel configurations. A limitation of NITAWL 

is that fuel rod subdivision it not permitted and fuel rods must be treated as a single region. 

The main difference between the depletion sequences is the neutron transport module. TRITON uses the 

2-D arbitary polygonal mesh discrete ordinates transport code NEWT to solve for the neutron flux 

distribution in the assembly. NEWT represents geometrical bodies as arbitarary polygons and uses the 

Extended Step Characteristic method to solve for the spatial discretization. NEWT uses an automated 

grid generation feature that makes it easy for the user to define complex configuration. The version of 

NEWT released in SCALE 5.1 uses a combinatorial geometry package, similar to KENO-VI geometry, 

that allows the user to easily migrate between 2-D deterministic and 3-D Monte Carlo models. 

The SAS2H depletion analysis sequence uses the 1-D discrete ordinates code XSDRNPM, which 

combines features of earlier ANISN, GAM-II, and THERMOS computer programs. To enable more 

accurate modeling of heterogeneous systems, XSDRNPM is implemented in SAS2H using a two-pass 

approach that first uses cell homogenization and then applies the homogenized cross sections in a second 

model that may include nonfuel water regions, absorber rods, or burnable poison rods in an assembly. 

The SAS2H methology and modeling procedures are described in more detail in Sect. 5.2. The NEWT 

geometry provides greater flexibility and accuracy in modeling heterogeneous lattices, and it allows 
75 

simulation of multiple fuel types (e.g., different enrichment zones, burnable absorbers, etc.) that can be 

independently depleted. 

After the transport calculation is complete, the neutron flux solution is used to weight the SCALE 

multigroup cross sections and prepare effective one-group cross sections for use with ORIGEN-S for all 

nuclides available in the library. The COUPLE code performs the function of updating the ORIGEN-S 

cross-section library with the weighted problem-dependent cross sections derived from the transport 

calculations. At defined burnup intervals during the simulation, the updated nuclide concentrations from 

ORIGEN-S are applied to the transport analysis of the assembly lattice, and the neutron flux spectrum is 

recalculated. The cross sections are again updated and used by ORIGEN-S for the following 

computational step. 

The typical calculational flow used in SCALE depletion sequences is shown in Figure 5.1 for a single 

burnup step. After this sequence of calculations is complete, the isotopic concentrations from the 

ORIGEN-S calculation are used to update each of the codes in the sequence to reflect the change in 

compositions and other reactor operating conditions with time. The sequence of calculations is repeated 

until the final discharge burnup is achieved. 

The decay heat calculations based on the TRITON depletion sequences can be performed in two ways. 

One approach is to perform the depletion and decay analysis for each individual assembly directly using 

TRITON. However, there is a significant computational time associated with using the depletion 

sequence, particularly for the resonance self-shielding and NEWT transport calculations. A second and 

more efficient approach consists of (1) performing TRITON depletion simulations for each assembly type 

to generate a burnup-dependent cross-section library for the assembly, and then (2) using the generated 

library to perform standalone ORIGEN-S irradiation and decay simulations for each assembly having the 

same, or similar design and operating characteristics. 

Time-dependent cross-section libraries are applied to standalone ORIGEN-S calculations using the 

ORIGEN-ARP sequence in SCALE. This sequence uses the Automatic Rapid Processing (ARP) module 

in SCALE to interpolate cross sections in an ORIGEN-S library parameterized as a function of fuel 

burnup, initial enrichment, and moderator density (in the case of BWR assemblies). Therefore, the user 

can employ the depletion sequences to generate burnup-dependent libraries that cover a range of discrete 

fuel parameters (e.g., enrichment and coolant density) and then use ARP to interpolate problem and timedependent cross sections for any fuel type within that range. In effect, the calculation is performed using 

the same cross sections as those used in the depletion sequence analysis, and the method produces results 

with the same accuracy. However, the calculations are performed in a small fraction of the time required 

by coupled depletion simulation because all resonance processing and transport calculations have been 

performed in advance and the cross sections saved for subsequent use. This methodology was employed 

as the reference method for all TRITON calculations presented in this report. A more detailed description 

of the procedures in given in Sect. 5.1.5. Calculations were also performed for selected assemblies using 

SAS2H. However, these calculations were performed directly using the SAS2H sequence (without crosssection interpolation) because the sequence runs relatively quickly. 
76 

Figure 5.1 Calculational flow of SCALE depletion analysis (1-D and 2-D) for one time step 

BONAMI

NITAWL

SCALE 

Multigroup 

Transport 

Library 

XSDRNPM 

(pin cell) 

XSDRNPM 

(1-D assembly) 

COUPLE

CENTRM / 

PMC 

NEWT 

(2-D assembly) 

ORIGEN-S 

cross section 

library 

(updated) 

ORIGEN-S

Resonance 

cross 

section 

processing 

Transport 

analysis 

Cross section 

library 

updating 

Depletion and 

decay 

analysis 

Continuous 

Energy 

Resonance 

Data 
77 

5.1.3 SCALE Cross-Section Libraries 

The transport calculations (with XSDRNPM or NEWT) can be performed using any of the SCALE 

multigroup neutron cross-section libraries. The libraries recommended for use with the depletion analysis 

sequences are the SCALE 44-group and 238-group libraries, both based on ENDF/B-V. These libraries 

contain evaluated cross section data for more than 220 nuclides that can be collapsed and applied in the 

ORIGEN-S burnup simulations. The 44-group library is a broad-group version of the 238-group library, 

collapsed using a fuel cell spectrum based on a typical 17 × 17 fuel assembly. All transport calculations 

in this study were performed using the SCALE 44-group ENDF/B-V cross-section library. Resonance 

cross section selfshielding was performed using the BONAMI and NITAWL modules. 

5.1.4 ORIGEN-S Data Libraries 

All nuclear decay and cross-section data in the ORIGEN-S libraries have been updated with modern data. 

Decay data are derived from the Evaluated Nuclear Data Files (ENDF/B-VI) and the Evaluated Nuclear 

Structure Data File (ENSDF). Cross-section data have also been entirely updated with modern 

evaluations to include a wide range of reaction types. The neutron intensity and energy spectra 

calculations performed by ORIGEN-S are based on methods developed in the SOURCES code,48

providing substantial improvements as compared with other codes, particularly for the neutron source 

associated with (α,n) reactions in the fuel matrix. In addition, all gamma ray intensity and spectra 

calculations are performed using an updated library of line-energy photon yields that can generate spectra 

in any energy group structure. 

Multigroup cross sections for nuclides in the SCALE cross-section libraries can be weighted with the flux 

spectrum from the transport calculation and then used to prepare the effective one-group cross sections 

for use in the ORIGEN-S calculation. The maximum number of nuclides that can be updated with this 

procedure is about 220; the number of isotopic evaluations within the ENDF/B-V libraries. However, the 

ORIGEN-S library presently has reaction data for 854 nuclides. ORIGEN-S cross sections for those 

nuclides not updated with data from either the 44- or 238-group SCALE library are obtained using a 

different procedure. In these cases cross sections from a standard broad-group cross-section library 

available with ORIGEN-S are employed. The broad-group (3-group) cross sections, derived from 

continuous-energy cross section evaluations weighted using a typical midlife fuel spectrum for a PWR, 

are collapsed using time-dependent flux weighting factors derived from the problem-dependent flux 

solution from the transport calculation. Cross sections are available for the following reaction types: 

fission, radiative capture (n,γ), (n,2n), (n,3n), (n,p) and (n,α). 

The ORIGEN-S libraries also contain direct fission product yields for 30 fissionable actinides: 

227,228,232Th, 231Pa, 232-238U, 238-242Pu, 241,242m,243Am, 237,238Np, 242-246,248Cm, 249,252Cf, and 254Es. The yields are 

derived from ENDF/B-VI. 

A detailed description of the sources of decay data and cross-section data implemented in the ORIGEN-S 

libraries can be found in Ref. 49. 

5.1.5 ORIGEN-S Cross-Section Library Generation 

The SCALE depletion analysis sequences perform full problem-dependent burnup simulations using 

design-specific information provided by the user. The depletion sequences function by iteratively 

performing transport calculations to update the cross sections and burnup simulations to update the 

nuclide concentrations. At defined intervals throughout an irradiation simulation, the transport code 

recalculates the cross sections to reflect the changes with burnup (time) in nuclide concentrations and 
78 

other reactor operating conditions. These coupled code systems are complex. They require detailed fuel 

design and operational information and require significant computer resources. The need for repeated 

transport calculations to recalculate the time-dependent cross sections also makes them relatively slow to 

execute. 

As described previously, the cross-section libraries generated by these sequences may be saved and used 

later in standalone ORIGEN-S simulations. The libraries can be used for depletion calculations involving 

fuels having the same or similar configurations without the need to recalculate the cross sections each 

time. When cross-section libraries are generated over a range of discrete fuel parameter values, the cross 

sections can be interpolated to provide accurate depletion capability for a wide range of potential fuel 

conditions. 

In ORIGEN-ARP, this interpolation function is performed using the ARP utility module. ARP reads the 

prepared libraries and interpolates a problem-dependent library for ORIGEN-S using the specified input. 

Cross sections can be interpolated as a function of burnup, enrichment, and coolant density value for 

situations where the conditions of the application are different than the conditions used to generate the 

libraries. The cross-section interpolation scheme of ARP uses a maximum of four cross-section data 

points with a third-order Lagrangian polynomial fit. The interpolation function will automatically search 

the array of independent variables to determine the appropriate points to use for the fit. The function will 

automatically adapt to perform a lower-order polynomial fit if fewer than four points are available. For 

two enrichment points, linear interpolation is performed. The interpolation is performed on all cross 

section data in a library, not just the cross sections for the most important actinides. Interpolation is 

performed for (1) the full transition matrix, (2) the neutron absorption cross section array, (3) the fission 

cross section array, and (4) the fission cross section multiplied by the neutrons per fission. 

The interpolation procedure eliminates the need to perform reactor physics transport calculations during 

the burnup simulation because the parameterized cross sections are developed in advance. Because the 

time to perform resonance self-shielding and the neutron transport calculation typically represents the 

majority of time in a depletion analysis calculation, the procedure runs in a small fraction of the time 

required by other methods, while maintaining the full accuracy of a coupled reactor physics calculation. 

5.1.5.1 Library Generation Procedures 

Cross-section libraries were generated for each of the assembly designs used in this study using the 2-D 

transport methods of the TRITON depletion sequence. Cross sections were generated as a function of 

burnup and fuel enrichment for PWR assemblies; a third parameter, coolant density, was added when 

generating BWR cross-section libraries. Examples of the input files for each type of assembly design are 

listed in Appendix B. The basic steps followed in creating the cross-section libraries are described below. 
79 

1. TRITON/NEWT Lattice physics models for each assembly design were first created (see 

Appendix B). The models include a 2-D description of the assemblies, including geometry, material 

compositions, and temperatures. 

2. Extra nuclides were added at trace initial concentrations (10−20 atoms/barn-cm) in the fuel 

composition description. This procedure ensures that the nuclides are included in the transport 

calculation as the concentrations increase during irradiation and that weighted cross sections for these 

nuclides are applied in the ORIGEN-S depletion simulation. This is done automatically in TRITON 

using the keyword addnux=3 as a parameter entry in the input file. 

3. Depletion calculations were performed using burnup steps of 3,000 MWd/MTU each, extending up to 

72,000 MWd/MTU. Each depletion calculation produced a library containing 24 sets of burnupdependent cross sections. An additional step is added at the beginning of irradiation with a short time 

(10−15 d) to generate cross sections for effectively unirradiated fuel. Each set of cross sections is 

referenced and accessed by its position in the library. 

4. The library was subsequently thinned to remove cross-section sets in the burnup range where the 

cross sections show only small variations with changes in burnup (e.g., generally in the regime of 

higher burnup). The library thinning was performed using the ARPLIB utility of SCALE. The 

thinning significantly decreased library memory requirements. The positions and corresponding 

burnup for each cross section set in the initial and final (thinned) library are listed in Table 5.1. 

An example of the input for ARPLIB is listed in Appendix B. 

5. The burnup calculations described above were repeated for discrete 235U initial enrichment values to 

cover the range of enrichments corresponding to the assemblies considered in this study. The 

enrichment steps used in this study were 1 wt % 235U (e.g., 2.0 and 3.0 wt % 235U). 

6. Water moderator density for BWR assemblies can change from typically about 0.7 g/cm3

 at the core 

inlet (bottom of assembly) to 0.2 g/cm3

 or lower at the top of the assembly due to voiding of the 

coolant within the flow channel caused by boiling. The voiding has the effect of hardening the 

neutron spectrum in the fuel (shifting the spectrum to higher energy) and therefore influences the 

value of the effective cross sections. To account for this effect in the current study, libraries for the 

BWR assemblies were generated as a function of the effective moderator density using discrete 

values of 0.1, 0.3, 0.5, 0.7, and 0.9 g/cm3

. Libraries for PWR assemblies were created using a single 

average value for the water moderator density because the density does not change significantly with 

axial height. 

The procedure described above was used to generate cross-section libraries for each assembly design. 
80 

Table 5.1 Cross-section positions and burnup values created in 

the SCALE libraries 

Initial 

positions 

Final 

positions 

Burnup 

(MWd/MTU) 

1 1 0a

2 2 1500

3 3 4500

4 4 7500

5 5 10500

6 6 13500

7 7 16500

8 19500 

9 22500 

10 25500 

11 28500 

12 8 31500

13 34500 

14 37500 

15 40500 

16 43500 

17 9 46500

18 49500 

19 52500 

20 55500 

21 10 58500

22 61500 

23 64500 

24 67500 

25 11 70500

a

 These bolded values represent the burnup of the 11 cross-section sets 

retained in the final cross-section library. 
81 

5.2 SAS2H Assembly Models

The reference method for calculating decay heat in this study was to use the 2-D depletion sequence 

TRITON because it allows a greater accuracy and flexibility in modeling the fuel assembly 

configurations, particularly for the more complex and heterogeneous BWR assembly designs. However, 

the SAS2H 1-D depletion sequence in SCALE is well established and has been widely used and 

demonstrated to be appropriate for a wide range of fuel assembly designs through an extensive 

benchmarking and validation program.3,35,50–52 The 1-D method also has a much better computational 

efficiency. 

To provide a comparison between 1-D and the more recently developed 2-D depletion methods in 

SCALE, calculations were performed for selected assemblies measured at the Swedish CLAB facility 

using the SAS2H depletion analysis sequence to demonstrate the applicability of the 1-D models. A 

general description of the assembly models developed for SAS2H is given in Section 5.2.1. 

SAS2H uses the 1-D discrete ordinates transport code XSDRNPM to simulate the lattice physics for the 

fuel rods and the assembly. This is accomplished by first performing a pin cell calculation to generate 

homogenized cell-weighted cross sections for the fuel, clad, and coolant. The homogenized pin cell 

mixture and cross sections are subsequently applied in a second transport calculation that can model 

heterogeneity in the assembly. This procedure is shown in Figure 5.2 for an illustrative example of eight 

fuel rods surrounding a central non-fuelled cell (×), e.g., water-filled guide tube, instrument tube, 

burnable absorber, or poison rod. The first XSDRNPM calculation (A) is performed for a single fuel rod 

and surrounding moderator associated with the simple pin cell to generate a cell-weighted mixture. The 

homogenized cross sections are then applied to a second XSDRNPM transport model of the assembly 

(B), illustrated in Figure 5.2 with the nonfuel cell centrally located in the assembly. The assembly model 

is represented in XSDRNPM using cylindrical geometry with equivalent region volumes. This procedure 

allows the user to accurately represent one level of heterogeneity in the model that would otherwise not 

be possible with a single 1-D transport calculation. 

5.2.1 Modeling Techniques 

For assemblies with multiple absorber rods or nonfuel lattice cells, the general procedure used in SAS2H 

is to model a single nonfuel cell in the center of the assembly and surround that cell with the volume of 

the homogenized fuel mixture that preserves the ratio of the nonfuel and fuel material in the assembly 

model. For example, an assembly with 25 nonfueled water cells (1 instrument tube and 24 guide tubes) 

would be modeled as a central water cell surrounded by 1/25 of the total number of actual fuel rods in the 

assembly. Therefore, these models physically represent only a fraction of the entire assembly. The 

XSDRNPM transport model applies a reflective outer boundary condition to closely preserve the 

neutronic environment of the actual assembly. 

For more complex assembly designs the limitations of a 1-D model may require approximations in the 

way the assembly model is constructed, particularly for assemblies having several levels of heterogeneity, 

e.g., an assembly containing both guide tubes and absorber rods. For some configurations these 

approximations can lead to several possible models that could be equally justifiable. In addition, the 

SAS2H sequence allows only one fuel mixture type. Therefore, assemblies containing fuel rods of 

different enrichment can not be simulated; the assembly-average enrichment is typically employed in this 

case. 

The approach, described above, has been shown to be highly reliable for PWR assembly configurations 

with only one heterogeneous cell type (e.g., guide tubes or burnable absorber rods). Nonfuel regions are 
82 

generally uniformly distributed in the assembly, rather than clustered, and are symmetrical, which allows 

for the accurate subdivision of the assembly in the model, i.e., a subassembly model. 

Figure 5.2 Pin-cell homogenization procedure used in SAS2H 1-D assembly models 

Designs for modern BWR assemblies are generally more complex and heterogeneous than PWR 

assemblies. They often involve enrichment zoning, employ integral burnable poisons, have either small 

or large water rods (channels), and include regions of boiling water and non-boiling water in the channels 

and outside of the assembly flow tube. The multiple levels of heterogeneity require approximations in a 

1-D model, and different approaches can yield different models for the same assembly. 

Murphy53 previously investigated the effect of different 1-D modeling approximations on the predicted 

decay heat for modern BWR assembly designs. A potential consequence of using a standard subassembly 

model for BWR assemblies is that the higher density (nonboiling) water in the channel region outside the 

assembly flow tube is placed outside of the subassembly model in order to preserve the volume fractions 

in the model. This, in effect, introduces higher density water from the outer channel moderator inside the 

assembly, increasing the level of moderation in the assembly as compared with the real situation where 

the channel water is located only outside of the boiling water region in the assembly. Alternate models 

studied by Murphy included a full-assembly model (no subdivision) that allowed a more accurate 

representation of the moderator regions but could not explicitly model burnable poison rods (these rods 

were smeared with the fuel rods in the assembly). The physical rationale for this model was that the 

influence of the moderator may be more important than that of the poison rods because the poisons are 

generally depleted early in the cycle, whereas the moderator influence is present throughout the irradation 

of the assembly. The study concluded that different modeling approximations yielded only small 

differences in the decay heat predictions; these differences resulting primarily from the actinide 

compositions that are more dependent on spectral changes in the assembly than most of the fission 

products. The differences in the models were more apparent at long decay times where actinide 

contribution to decay heat becomes increasingly important. No single model prescription was found to be 

applicable to all designs. For example, BWR assembly designs such as the General Electric and 

(A) 

(B)
83 

Framatome ANP ATRIUM (see Figure 5.3) assembly designs, characterized by large central water 

regions, should not be subdivided (by the number of absorber rods) in a way that does not preserve the 

moderating effects of the internal channel(s). For these designs, preserving the moderating effects of the 

channel is important, and full-assembly models (without subdivision) typically yield more accurate 

results. For other designs utilizing both small water tubes and burnable poison rods, the subassembly 

model (for example, see Ref. 40) has been shown to yield accurate isotopic results. 

Figure 5.3 Cutaway illustration of the ATRIUM-10 

assembly layout with large internal square water channel 

(Source: AREVA Web site) 

5.2.2 BWR Swedish Assemblies 

5.2.2.1 8 × 8 Assemblies 

Most of the Swedish BWR 8 × 8 assemblies measured at CLAB used a small number of fuel rods 

containing Gd2O3 burnable absorber (BA). The conventional approach for modeling these assemblies40

has been to include a gadolinium-bearing rod in the center of the model and place the proportional 

volume of fuel around the rod. For the case of an 8 × 8 assembly that has five BA rods, for instance, the 

model represents one fifth of the assembly by putting one gadolinium-bearing rod at the center with 
84 

one-fifth of the fuel rods (e.g., one-fifth of the remaining 59 rods) surrounding this as the fuel area. 

The additional water in the lattice associated with any non-fuelled cells and water surrounding the 

assembly (inside and outside the assembly flow tube) is typically placed outside the subassembly fuel 

zone. 

A small modification was required because many of the Swedish 8 × 8 assemblies contained corner rods 

(three rods at each corner) having a smaller diameter than non-corner fuel rods. In these assemblies, the 

non-corner rods were used to define the pin cell dimensions. With the larger fuel rod diameter, the 

effective number of fuel rods in the assembly is adjusted such that the correct fuel mass (or, alternatively, 

fuel area) is preserved. This effective number of fuel rods is then used in determining the size of the fuelzone material in the approach described above. For the case of 64 fuel rods, twelve of which are corner 

rods, this equivalent number of normal-sized rods is roughly 62. 

The 8 × 8 assembly model includes a square Zircaloy box defining the outside of the assembly with 

channel moderator as the outermost region. In all cases, these materials were apportioned to preserve the 

correct fraction of the assembly being modeled. Furthermore, one must account for the small amount of 

moderator immediately to the inside of the Zircaloy box; the full compliment of 64 cells does not fill the 

entire area inside the box and the extra moderator must be taken into account. Again, the fraction that 

corresponds to the model of the assembly is included. 

The majority of the 8 × 8 assemblies contained one water rod, however, a small number contained none, 

and one assembly had two water rods. The few assemblies that had no BA rods and one water rod were 

each modeled as a full assembly surrounding the central water hole. The assembly with two water rods 

had uniform sized fuel rods (no corner rods) and four BA rods; it was modeled as one quarter 

subassembly model surrounding one BA rod. 

5.2.2.2 9 × 9 Assemblies 

Three Swedish 9 × 9 BWR assembly designs were measured at CLAB. All three contained five water 

rods and six BA rods per assembly; all fuel rods had a uniform diameter (i.e., no corner rods). The 

assemblies were modeled as one sixth of an assembly with the BA rod at the center, in a similar manner 

to that described for the 8 × 8 assemblies. 

5.2.2.3 SVEA Assemblies 

Two types of SVEA assemblies were studied: SVEA-64 and the SVEA-100. These are 8 × 8 and 10 × 10 

fuel lattices, respectively. As previously described, a unique design feature of the SVEA assemblies is a 

large water cross that subdivides the assembly into quadrants, or subassemblies (see Figures 4.8 and 4.9). 

The arrangement is composed of four 4 × 4 and four 5 × 5 sub-assemblies, respectively. All SVEA 

assemblies were modeled as one quarter of the full assembly, allowing the moderator in the water cross 

region to be accurately modeled around each quadrant, thereby preserving the moderating influence of the 

water cross region. 

Some assemblies contain one gadolinium-bearing (BA) rod per quadrant, which can be simulated 

explicitly in the SAS2H quadrant model. However, in assemblies having more than one BA rod per 

quadrant, the nominal amount of Gd2O3 in the central BA rod was adjusted by the factor N/4, where N is 

the total number of BA rods in the assembly. This approximation preserved the specified total amount of 

burnable poison in the assembly model and preserved the neutronic effects of the water cross region. The 

subassembly (quadrant) model then includes the central BA rod; the clad and water for the BA rod unit 

cell; the homogenized fuel zone having an area corresponding to the remaining number of fuel pins per 

quadrant; a water zone accounting for the water inside the Zircaloy box that encloses the quarter 
85 

assembly; the flow box for the quadrant; and finally the water associated with the water cross region. 

Note that all water contained in the box has a density that accounts for the void fraction. The zone 

outside the quarter assembly flow box accounts for water in the water cross between subassemblies and in 

the channel between assemblies (i.e., two sides of the quarter assembly are the water cross and two sides 

are the water between adjacent assemblies). The volume of the outer water zone is determined from the 

assembly geometry data. Any control-blade insertion for the assembly would be modeled by adding the 

absorber to the outermost water region, as described in previous studies.40

5.3 TRITON Assembly Models

The assembly design information in Sect. 3 was used to prepare detailed geometry models for each 

assembly design. The models were used in TRITON to prepare cross-section libraries for each type of 

assembly. These libraries were subsequently used for the ORIGEN-S decay heat calculations. Typical 

configurations, obtained from literature, were used to model assemblies for which configuration details 

were not available. Because many assemblies for a given design type (i.e., 8 × 8) are unique, it was 

important to establish which design characteristics were required to prepare cross-section libraries that are 

applicable to a broad range of configurations. For example, the 8 × 8 design included assemblies with 

zero to six burnable poison rods. Because the fuel cross sections are influenced by the number of rods 

and the poison content, it was necessary to determine the minimum number of unique libraries that were 

needed to represent the different configurations and determine which configurations could be grouped 

together for the purposes of predicting the decay heat. 

This section describes the procedure for generating burnup-dependent cross-section libraries for use with 

ORIGEN-S. It also discusses the additional assumptions made when developing the TRITON models for 

the assemblies considered in the present study. 

5.3.1 ORIGEN-S Cross-Section Libraries for U.S. Assemblies 

5.3.1.1 BWR GE 7 × 7 Library 

The assembly specifications were taken from Table 4.10 for the GE-3 Cooper fuel design data. 

The channel water density of 0.7627 g/cm3

 was applied, a value used in isotopic benchmark analyses 

documented in Ref. 40. It was noted that this value was slightly different than the value of 0.669 g/cm3

used in a previous benchmark analysis of the decay heat for the Cooper assemblies, reported in Ref. 3; 

this lower density value was derived to account for the displacement of water by the control cruciform by 

reducing the water volume of bottom axial node of the assembly by 10%. 

The GE 7 × 7 assemblies in this study used integral burnable gadolinium (Gd) poison rods with Gd2O3 as 

burnable absorber. The previous model used for Cooper station assembly decay heat analyses3

 assumed 

four Gd rods per assembly; a value based on typical BWR design information.54 The Gd rods 

arrangement used for generating the 7 × 7 cross-section libraries in the current study are based on the 

actual Gd rod configuration of Cooper assembly CZ346 that is documented in Ref. 36. This assembly is 

one of the assemblies measured in the decay heat program at Morris Operations. CZ346 had five Gd rods 

with an average Gd2O3 poison concentration of 3.4 wt % (three rods with 3.0 wt %, and two rods with 

4.0 wt %). The assembly model illustrating the Gd rod configuration is shown in Figure 4.4. The crosssection library generated for this assembly configuration was used for the analysis of all 7 × 7 assemblies, 

including the Dresden and Monticello assemblies. 
86 

5.3.1.2 PWR 14 × 14 Library 

PWR 14 × 14 libraries were generated to calculate decay heat for 14 × 14 assemblies from San Onofre 

Unit 1 reactor and Point Beach Unit 2 reactor. The design differences between the two types of 

assemblies are significant, as can be seen from the assembly data presented in Tables 4.2 and 4.8. 

Therefore, two different cross-section libraries were generated, one for each type of reactor. One fuel 

enrichment value, of 3.397 wt % 235U, was considered for the purpose of generating burnup-dependent 

cross sections for Point Beach assemblies, as all of the measured assemblies had the same enrichment; for 

San Onofre assemblies two fuel enrichment values were used, 3.5 and 4.0 wt % 235U, to span the range of 

enrichments for all analyzed assemblies from this reactor. The TRITON assembly models are presented 

in Figures 4.1 and 4.3. 

In terms of the predicted decay heat, a significant design difference between the San Onofre and 

Point Beach reactor assemblies is the use of stainless steel cladding for the San Onofre assemblies. 

The activation of trace levels of cobalt in steel represents a significant source of decay heat for these 

assemblies. To accurately simulate the activation of 59Co in the clad, cell-weighted cross sections are 

needed in the ORIGEN-S library that reflect the differences in the neutron flux level and energy spectrum 

between the fuel and the clad. This was done in the TRITON depletion calculations by defining separate 

depletion materials in the problem for the fuel and the clad regions. A combined cell-weighted crosssection library is automatically generated by TRITON. The weighing procedure adjusts the cross sections 

in the combined library to reflect the flux level and spectral differences between the different regions in 

which each nuclide resides in such a way as to preserve the reaction rates in an homogenized, or point 

model. 

5.3.1.3 PWR 15 × 15 Library 

The library for the 15 × 15 assembly was generated using the design data presented in Table 4.5 for the 

Turkey Point 3 reactor. All measured assemblies have the same fuel enrichment, 2.557 wt % 235U; 

therefore, the burnup-dependent cross sections were generated only at this particular enrichment value. 

The configuration of the TRITON model is shown in Figure 4.2. 

5.3.2 ORIGEN-S Cross-Section Libraries for Swedish Assemblies 

Significantly more detailed fuel design and reactor operating information was available for the Swedish 

fuel assemblies measured at CLAB than was available for the U.S. measured assemblies. The design data 

for the Swedish fuel included the number of burnable poison rods and the concentration of the burnable 

poison in each assembly. Axial void profiles were provided for most BWR assemblies, as well as axial 

burnup profiles provided by the utilities. More detailed information can be found in Ref. 39. 

5.3.2.1 BWR 8 × 8 Libraries 

TRITON models used to generate the ORIGEN-S cross-section libraries are developed based on the 

assembly fuel design data provided in Table 4.15. Four types of assembly design are considered, which 

show slight variations in dimensions, number of burnable absorber (BA) rods and concentration of the 

absorber material in the absorber rods. Only one TRITON model was created for Types 1 and 5 because 

these assemblies had similar characteristics. For Type 2 and Type 4 however, multiple TRITON models 

were used because of differences in the number of BA rods and/or the concentration of the burnable 

absorber. The configuration of the assemblies with no BA rods is shown in Figure 4.6. 

For Type 2 assemblies, the number of BA rods can be zero, three, or four. For the cases where 

three BA rods are present, the concentration of the absorber material (Gd2O3) in the BA rods can be 
87 

3.95 or 5.5 wt %. A separate cross-section library was generated for each of these 4 groups of Type 2 

assemblies. Because information was not available on the BA rod locations in the assembly, assumptions 

were necessary. The placement of the BA rods in the assembly is illustrated in Figure 5.4 for assemblies 

with 3 and 4 BA rods, respectively. 

Figure 5.4 Layout for Type 2 assemblies with three and four BA rods 

In the Type 4 assembly design there are assemblies with no BA rods and with three, four, or five BA rods 

having a Gd2O3 wt % of 2, 2.55, and 3.2, respectively. The same TRITON model was used to generate 

cross-section libraries for the all 8 × 8 assemblies with BA rods. An assembly with configurations of 3 

and 4 BA rods is shown in Figure 5.4. The arrangement of the BA rods for Type 5 assemblies is similar 

to that shown in Figure 5.4; however, no corner rods are present. 

Most of the 8 × 8 assemblies have the initial 235U enrichment in the range 2–3 wt %, except the Type 4 

assemblies with five BA rods. The cross-section libraries were generated for enrichment values of 2 wt % 

and 3 wt % 235U for all assemblies categories but Type 4 with five BA rods, for which an additional 

4 wt % value was considered. The same set of water moderator density values (0.1, 0.3, 0.5, 0.7, and 

0.9 g/cm3

) for which the cross section are obtained was used in all cases. 

5.3.2.2 BWR 9 × 9 Libraries 

The TRITON model for the 9 × 9 assemblies, based on the design data shown in Table 4.24 is illustrated 

in Figure 4.7. Because all the assemblies in this group have the same initial fuel enrichment of 

2.938 wt % 235U, burnup-dependent cross sections are generated only for this enrichment value and at five 

coolant density points as specified previously for the 8 × 8 assemblies. 
88 

5.3.2.3 BWR SVEA Libraries 

The burnup-dependent cross-section libraries for the SVEA-64 and SVEA-100 assemblies were generated 

at five fuel enrichment values (1.5, 2.0, 3.0, 4.0, 5.0, and 6.0 wt % 235U) and the same set of coolant 

density points as for all other BWR assemblies studied. The assembly design data applied in the 

TRITON models are given in Tables 4.25 and 4.28, and the assembly configurations are illustrated in 

Figures 4.8 and 4.9. Even though there are slight differences in the actual number of BA rods and/or the 

Gd2O3 content, only one representative TRITON model was used for each SVEA design: four BA rods at 

2.55 wt % Gd2O3 were used for the SVEA-64 model and six BA rods at 3.5 wt % Gd2O3 for the SVEA100 model. 

5.3.2.4 PWR 15 × 15 Libraries 

The assembly design data used in the TRITON models is presented in Table 4.32. BA rods were not used 

in any of the 15 × 15 assemblies measured at CLAB. Two types of assemblies are shown in the table, 

differing by the dimension of the pellet diameter, rod diameter, and clad thickness. A separate TRITON 

model was used to generate a library for each type. The assembly layout is illustrated in Figure 4.10. 

The burnup-dependent cross sections were generated for only one fuel enrichment value, of 3.095 wt % 

235U, for assemblies of Type 1, as all the assemblies of this type had the same enrichment; for assemblies 

of Type 2, two enrichment values were used, 3.0 and 3.5 wt % 235U, to cover the range corresponding to 

all assemblies of this type. 

5.3.2.5 PWR 17 × 17 Libraries 

Two models were created for the 17 × 17 assemblies: one with no BA rods and one with 12 BA rods. 

The detailed design information was presented in Table 4.35, and data on the number of BA rods for each 

assembly studied in Table 4.36. The configuration of the assembly containing 12 BA rods, based on 

information in Ref. 41, is illustrated in Figure 4.11. For both TRITON models, libraries were generated 

at four fuel enrichment values (2.0, 2.5, 3.0, and 3.5 wt % 235U). The coolant density for all PWR 

assemblies (including 15 × 15) was 0.72 g/cm3

, a value appropriate to the actual pressure and coolant 

average temperature values cited in Ref. 55 for each reactor. For example, the coolant data for 17 × 17 

assemblies are 15.25 MPa pressure and a temperature of 576 K; the corresponding density can be 

determined from standard pressure-temperature tables. 
89 

6. COMPARISON OF PREDICTIONS AND MEASUREMENTS 

This section compares predicted and measured decay heat rates for the assemblies measured at 

GE-Morris, HEDL, and the Swedish CLAB facilities. The calculations were performed using reactor 

operating data and fuel assembly specifications described in Sect. 4 and the SCALE computational 

methods and software tools presented in Sect. 5. All decay heat calculations were performed with the 

ORIGEN-S code using cross sections interpolated to the corresponding initial fuel enrichment, irradiation 

history, and operating conditions using the ARP module. Cross-section libraries for the ORIGEN-S 

depletion analysis were generated with the 2-D TRITON depletion sequence. Other results presented in 

this section include calculations that account for the axial variation of the burnup profile and void profile 

(in the case of BWRs only) for the Swedish assemblies that had detailed axial data available. 

Comparisons using the 1-D SAS2H depletion sequence are also described for selected Swedish 

assemblies. Finally, an estimate of the experimental error and the bias and uncertainty associated with the 

calculations is presented based on the comparisons of the calculations with experimental data. 

6.1 Assembly Decay Heat Results

Comparisons of the measured decay heat rates for all evaluated assemblies with the corresponding values 

predicted by ORIGEN-S are listed in the tables of Appendix A. The results are plotted on a log-log scale 

(to clearly display large number of data points for low values of decay heat) in Figure 6.1. The results for 

the 198 assemblies are organized in the tables by measurement facility, reactor, and assembly design. 

Repeat measurements are indicated in the tables. 

The GE-Morris measurements presented in this report are those obtained from operating the calorimeter 

in static mode. The recirculation mode measurements reported for some assemblies exhibited larger 

variability than the static mode results and were therefore not used in this study. 

The Monticello assembly measurements, performed at GE-Morris Operations, were complicated due to 

difficulties in calorimeter calibration. Two separate calibrations, one performed in 1984 and the other in 

1985, resulted in significantly different decay heat results for the same assemblies. The 1985 calibration 

yielded a reproducibility of about ± 2 W, whereas the 1984 calibration was no better than ± 15 W based 

on repeat measurements of selected assemblies.30 However, an analysis of the measurements by 

McKinnon et al. indicated that the results using the 1985 calibration method yielded a 20-W average bias 

between measurements and predictions, whereas the 1984 calibration method resulted in less than a 4-W 

average bias. The cause of the differences, either in the method of calibration or operation, was never 

fully resolved. In this study the measurements reported using the 1984 calibration were used for the 

reason that the 20-W average bias observed with the 1985 data has not been observed in any of the other 

experiments. 

The error in the residuals (the absolute difference between the calculations and measurements) is a useful 

measure of calculational and measurement error. The residuals, in units of Watts, are plotted against the 

measured decay heat in Figure 6.2. The relative error is the residual error divided by the measured decay 

heat. Plotting the relative percent error vs decay heat for all assembly measurements produces the 

distribution shown in Figure 6.3. 
90 

10 100 1000

10

100

1000

 CLAB

 GEMorris

 HEDL

Predicted decay heat (W)

Measured decay heat (W)

Figure 6.1 Measured vs predicted decay heat for all assemblies. The solid line 

represents the line of agreement between predictions and measurements 

(i.e., slope = 1). 
91 

-60

-40

-20

0

20

40

60

0 200 400 600 800 1000 1200 1400 1600

Measured Decay Heat (W)

Residual Error (W)

Figure 6.2 Absolute error between calculated and measured decay heat 

-80

-60

-40

-20

0

20

40

0 200 400 600 800 1000 1200 1400 1600

Measured Decay Heat (W)

Relative Error (%)

Figure 6.3 Relative error between calculated and measured decay heat 
92 

6.2 Error and Uncertainty Analysis 

The results of the ORIGEN-S calculations are compared with the measured decay heat for the spent fuel 

assemblies to assess the accuracy of the calculations and the uncertainty in the calorimeter measurements. 

The objective of the analysis is to estimate the uncertainty that should be assigned to spent fuel assembly 

decay heat predictions made using the ORIGEN-S code. 

Error is introduced into the calculations through a variety of sources including (1) the numerical method 

and its implementation, (2) the nuclear cross sections and decay data, (3) errors in the input data 

(e.g., operational data, burnup accuracy, etc.), and (4) simplifying assumptions in the modeling 

(e.g., operating history, assembly design, axial variations). Errors are also associated with the calorimeter 

measurements due to experimental uncertainties (accounting for escaped radiation, heat capacities of the 

assembly, etc.) as well as the measurements of quantities such as temperature and time. 

6.2.1 Statistics for Uncertainty Analysis 

In evaluating the calorimetric data and the calculations, it is useful to make a variety of comparisons, as 

will be done in succeeding sections. However, in obtaining a quantitative uncertainty measure, it is 

important to select the statistical variables carefully. In general, we seek to compare a calculated quantity 

x to a quantity that can be measured or derived based on a measurement y. The error in both 

measurement and calculation is represented with an additive error term ε, 

 y = x + ε . (1) 

If we have N pairs of experimental and calculational results (xi, yi), i = 1, 2, . . ., N, then evaluating the 

resulting error terms can give us a good understanding of the combined uncertainty between the 

calculation and the measurements. If the errors are all independent samples from a normal distribution 

with mean μ and variance σ

2

, then these two quantities can be approximated by the respective sample 

statistics: 

ε = iΣ εi / N (sample mean) 

 s2

 = iΣ (εi – ε )

2

 / (N – 1) (sample variance). 

If the mean μ = 0, then the distribution is unbiased, a highly desirable result. In general, the actual mean 

and variance are not known, and so the sample mean and variance are the only statistics we have 

available. Hence, it is important to ensure that we have valid estimates for these sample statistics, which 

may place some constraints on exactly which quantities x and y we should actually consider. Therefore, 

we describe the assumptions that will enable such analysis: 

1. The errors must be independent. We assume that all calorimetric measurements are independent. 

2. The errors all have the same normal distribution. That is, the variance σ

2

 is constant, and the 

collection εi, i = 1, 2, . . ., N represent random samples of a normal distribution. 

In evaluating the data, three possible approaches to the uncertainty analysis were considered: 

Decay heat error residuals. The error residuals are simply the difference between the calculated and 

measured decay heats, so that x and y in Eq. (1) are the calculated and measured decay heats, and ε has 

units of Watts. As will be shown, this is a reasonable choice for the present set of data, since the error of 
93 

the residuals has a variance that is nearly constant (see Figure 6.2) and is almost normally distributed with 

a sample mean near zero. These observations will be further discussed in a later section, and the 

uncertainty analysis developed using this statistical approach. 

Decay Heat Relative Error. In the case, y = (C/E – 1)%, where C/E is the ratio of calculated-toexperiment decay heat, the error is dimensionless. An evaluation of the relative error for all results 

indicates this is not a good measurement of error, because Figure 6.3 shows that the relative error 

decreases as decay heat increases, i.e., σ

2

 is not constant. Additional analyses for the data also indicate 

the distribution is not normal. The relative error can be a useful quantity for qualitative error analysis, but 

is not appropriate for uncertainty analysis of the data in this study. 

Specific Decay Heat Error Residuals. For both calculation and measurements, the specific decay heat 

was evaluated using the decay heat for the assembly divided by the initial uranium loading. The error 

residuals are then the difference between calculated and measured specific decay heats with units of 

W/MTU. This quantity was evaluated in comparing results from assemblies of different sizes 

(e.g., BWRs and PWRs) that have significant difference in the initial uranium loadings. However, 

dividing by the initial loading (effectively weighting by the inverse of the loading) produced error 

residuals using the present data that are not normally distributed. The specific decay heat is used in this 

analysis for illustrative purposes only, but not to calculate uncertainties. 

6.2.2 Calorimeter Measurement Error 

In assigning error to the calorimeter data, the measurements were evaluated independently of the 

calculations by making use of multiple measurements that were available for several assemblies. 

While some assemblies have only a single measurement, there are a number that have more than one 

measurement; some just a few days apart, others spanning months or even years. We evaluated these 

multiple measurements as representative of all assemblies. This approach is viable since most of the 

measurement error is characteristic of the calorimeter design, calibration, operation, and analysis, and is 

not directly related to the characteristics of individual assemblies. We assumed that all calorimeter 

measurements, even those on the same assembly, are independent data points. This is justifiable, since in 

each case, the assembly is loaded, sealed, and evaluated independently of any other measurements that 

have been done. 

If two or more measurements on the same assembly (having decayed for at least several years) are made 

only days apart, the results should be virtually identical, since very little decay occurs during this time. 

In this case, measurement uncertainty can be approximated by the sample standard deviation s, which is 

the square root of the sample variance: 

 

( ) ( ) 2 2

/ 1, is x xn =− − ∑ i (2) 

where x = i

Σ xi/n is the sample mean, n is the number of multiple measurements on that assembly, and xi

are the decay heat measurement results. If several assemblies have such closely-spaced multiple 

measurements, then the overall sample variance is obtained by combining the variances for individual 

assemblies: 

 

( ) ( ) 2 2

tot 1 / 1, k kk k k

s n sn =Σ − Σ − ⎡ ⎤ ⎣ ⎦ (3) 

where the index k has been used to denote the individual assemblies, and nk are the number of 

measurements for assembly k. 
94 

If multiple measurements span many weeks or months, then consideration must be made for the decrease 

in decay heat over that time period. Such a change may be approximated by a straight line, in which case 

the correlation between measurements is defined by a regression curve. Recognizing that decrease may 

not be linear, but rather exponential, it is also possible to fit an exponential function to the data. In either 

case, uncertainty is estimated with the sample standard deviation. However, a linear or exponential fit of 

the variance is determined not from the deviation of replicate points from each other, but by the deviation 

of each point from the best-fit curve. In the case of linear regression, the sample variance takes the 

form:56 

s

2

 = iΣ (a ti + b – xi)

2

 / (n – 2) , (4) 

where again xi are the measured decay heat values and ti are the times at which they are measured. 

The constants a and b are the slope and intercept, respectively, of the regression line. The sample 

variance is determined by dividing the sum of squares by n – 2, since two statistics (a and b) are to be 

determined from the n independent measurements. For this reason, at least three data points are necessary 

to make an estimate of variance for such a correlated series of measurements. 

Assemblies with multiple measurements appropriate for calorimeter uncertainty analysis are indicated in 

Table 6.1 and the uncertainty results for each facility is summarized. Regression analysis (measurements 

over an extended time interval) was performed for three assemblies: CZ205 and CZ259 (measured at 

GE-Morris), and D-15 (measured at HEDL). For the first two assemblies, a straight line approximation 

was equivalent to an exponential fit, and so this was used for simplicity. In these cases, measurement 

error was much greater than data variations, and so the form of the fitting function was of minor 

importance. For the HEDL assembly measurements, an exponential function was used to fit the data. 

The data for these three assemblies are shown in Figure 6.4, together with the regression curves. Note 

that there is considerable scatter in both sets of data measured at GE-Morris. 

All other replicate measurements were close enough in time to justify analysis based on the sample mean 

of the measurements. In this case, the greatest time spans were for Monticello assembly MT133 

(14 days) and CLAB assembly 6432 (18 days). The predicted decay heat decreases over this time by 

about 0.2 W; the variations in calorimeter measurements far exceed this change. Thus, the uncertainty 

approach based on the sample mean [Eq. (2) assuming constant decay heat] is reasonable in these cases. 

Shown in Figure 6.5 are four assemblies with replicate measurements from the GE-Morris facility. 

In each case, discernable scatter from the mean exists. On the other hand, the CLAB data exhibit much 

less scatter so that replicate points are usually very close to each other, as observed in Figure 6.6. 
95 

Table 6.1 Measurement uncertainty estimates for each calorimeter facility 

Measurement 

facility Assembly 

Number of 

repeat 

measurements 

n 

Time range of 

repeat 

measurements 

(days) 

Analysis 

methoda

Absolute 

standard 

deviation 

(W)

GE-Morris DN212 2 15 m 5.9 

GE-Morris CZ205 14 246 r 14.1 

GE-Morris CZ246 2 3 m 10.4 

GE-Morris CZ259 3 197 r 30.9 

GE-Morris MT123 3 6 m 16.7 

GE-Morris MT133 4 14 m 22.6 

GE-Morris MT228 3 12 m 15.9 

Total uncertainty 16.7 

 

HEDL D-15 3 1115 r 95.1 

CLAB 6432 2 2 m 2.00 

CLAB 6432 5 18 m 1.55 

CLAB 3838 2 1 m 0.45 

CLAB 9329 2 2 m 0.80 

CLAB 6350 2 1 m 1.77 

CLAB 13847 2 0 m 0.35 

CLAB C20 4 9 m 8.24 

CLAB E38 2 1 m 0.99 

CLAB 5A3 3 2 m 3.64 

CLAB 5A3 2 13 m 0.34 

Total uncertainty 3.94 

a

 Method of calculating standard deviation, s: r = regression, m = mean. 
96 

240

260

280

300

320

340

360

800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900 2000 2100

Decay Time (d)

Decay Heat (W)

400

600

800

1000

1200

1400

Decay Heat (W)

CZ205-data

CZ205-line

CZ259-data

CZ259-line

HEDL-data

HEDL-curve

Figure 6.4 Regression analysis of repeat measurements 

0

20

40

60

80

100

120

140

160

180

2500 3000 3500 4000 4500

Decay Time (d)

Decay Heat (W)

DN212

MT123

MT133

MT228

Figure 6.5 Repeat measurement results at the GE-Morris facility 
97 

100

200

300

400

4000 5000 6000 7000 8000

Decay Time (d)

Decay Heat (W)

6432-a 6432-b

3838 9329

6350 13847

C20 E38

5A3-a 5A3-b

Figure 6.6 Repeat measurements made at the Swedish CLAB facility 

CLAB assembly 6432 involved measurements more than 8 months apart; however, they occur in two 

clusters of cooling time with no measurements in between. Each cluster is therefore evaluated separately, 

and calculation of uncertainty is based on the sample mean for each cluster. If all points were treated 

using regression, the slope and intercept of the regression line would largely reflect the time difference of 

these two clusters (like fitting a line to 2 points), and would contribute little additional information to the 

uncertainty analysis. A similar situation exists for CLAB assembly 5A3, which has two clusters of 

measurements almost 1 year apart. 

For the GE-Morris calorimeter, the uncertainty based on repeat measurements for individual assemblies 

range from 5.9 to 30.9 W, and the overall value of 16.7 W, obtained from Eq. (2), is of similar magnitude 

to the values for the individual assemblies. For the HEDL calorimeter, only one assembly had multiple 

measurements. The uncertainty of 95.1 W is considerably larger than the uncertainties for GE-Morris or 

CLAB. However, the HEDL measurements involved assemblies with much higher decay heat, so one 

might expect larger absolute uncertainties. The results for the CLAB measurements indicate a lower 

uncertainty of 3.9 W. 

6.2.3 Differences Between Calculations and Experimental Data 

Having evaluated the uncertainties inherent in the measured decay heat data, the data were compared with 

the prediction. Initial analysis was performed using the error residuals 

εi = yi – xi , (5) 

where yi are the calculated decay heat values, and xi are the measured values. From the plot of the 

residuals in Figure 6.2 it can be seen that the calculation errors are for the most part scattered within 20 W 

of the measurements. 
98 

Noteworthy is a group of negative residuals exceeding 20 W in the measured decay heat range of 

300–400 W. Except for this group, the residuals appear equally distributed above and below the line of 

zero error, indicating that the calculations are as likely to overpredict as underpredict relative to 

measurement. The mean of the residual errors ε = –5.03 W for all measured assemblies combined. 

This error is similar to the standard deviation associated with the CLAB measurements (3.9 W) and well 

within those of GE-Morris (16.7 W) and HEDL (95 W). The standard deviation of the residual error 

about this mean is sε = 16.14 W, similar to the GE-Morris facility measurement uncertainty, substantially 

lower than that of the HEDL facility, and somewhat larger than the CLAB facility uncertainty. 

Except for the cluster of negative residuals noted above, the residual error has a distribution that is near 

normal, as illustrated by the approximate probability density function (PDF) in Figure 6.7. This PDF was 

derived from a smoothed histogram of the residual error. The principal peak is very near zero, and the 

right tail of the distribution is consistent with a normal distribution. However, for the left-hand tail, there 

is a second (smaller) peak that derives entirely from the aforementioned cluster of negative residuals. 

Further examination of this second peak shows that the data are for BWR assemblies measured at 

GE-Morris. The presence of bias in these measurements will be discussed at greater length later. 

In spite of the bimodal form of the PDF, the entire distribution of residuals is still approximately 

distributed normally. A comparison of the two cumulative distribution functions is shown in Figure 6.8 

(the residual distribution has been standardized as shown in the table). Using the Kolmogorov-Smirnov 

method57 to make a quantitative comparison, the Kolmogorov-Smirnov statistic is D = 0.1256 (the 

maximum vertical difference between the two cumulative distributions). With a sample size of 199 

points, the resulting P-value is 0.08, which suggests that the hypothesis of normality cannot be rejected 

with 95% confidence, but it can with 90% confidence; hence, the distribution might be normal, but is by 

no means assured. 

This preceding analysis treats all error the same, regardless of the source. Further evaluation of the data 

was performed to investigate the presence of other correlations and biases. Shown in Figure 6.9 is the 

error from the three different measurement facilities, which indicates several biases. The GE-Morris 

errors exhibit a negative bias, whereas the CLAB and HEDL errors are biased slightly positive 

(e.g., calculations are somewhat larger than measurements). This effect is quantified by the mean values 

of residual error due to the different measurement facilities: ε GE = −11.5 W, ε HEDL = 14.6 W, ε CLAB = 

2.1 W. The causes of these discrepancies are not obvious, but are likely related to differences in either 

experimental procedures or input data used in the calculations (or both). An important finding is that the 

mean values for each facility are within the data uncertainties identified in Sect. 3. 

From Figure 6.9 there also appears to be a slight trend in residual error for the HEDL measurements, 

although the correlation is not strong due to the large scatter in the data. Such a trend is noticeably absent 

for the GE-Morris and CLAB data, although both show significant data clustering, which bears further 

scrutiny. 

The two data clusters in the CLAB data shown in Figure 6.9 can be generally grouped by reactor type, as 

illustrated in Figure 6.10. The BWR data have smaller error residuals and are centered approximately 

about zero; they also indicate a small decrease in absolute error with increasing decay heat, although it is 

difficult to quantify because of the random scatter of the data. By contrast, the PWR data indicate a clear 

positive bias. It is quite likely that there is either systematic error in the calorimeter measurements for the 

PWR assemblies or in the calculations (nuclear data or input data), which causes this bias. However, the 

effect is small, and for this reason it may be difficult to identify. 
99 

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

-3.5 -3 -2.5 -2 -1.5 -1 -0.5 0 0.5 1 1.5 2 2.5 3 3.5

Random Variable (in Multiple of Standard Deviation σ)

Probability

Figure 6.7 Approximate probability density function for error residuals 

(Ordinate axis represents units of sample standard deviation, i.e., 

0sε, ±1sε, ±2sε, ±3sε) 

Figure 6.8 Comparison of standardized sample and normal cumulative 

distributions (Ordinate axis represents units of sample standard deviation, i.e., 

0sε, ±1sε, ±2sε, ±3sε) 

0

0.2

0.4

0.6

0.8

1

-3 -2 -1 0 1 2 3

Sample

Normal
100 

-60

-40

-20

0

20

40

60

0 200 400 600 800 1000 1200 1400 1600

Measured Decay Heat (W)

Residual Error (W)

GE-Morris

CLAB

HEDL

Figure 6.9 Residual error for each measurement facility 

-10

-5

0

5

10

15

20

0 100 200 300 400 500 600 700 800

Measured Decay Heat (W)

Residual Error (W)

BWR

PWR

Figure 6.10 Residual error for CLAB measurements by reactor type 
101 

The clustering of residuals by reactor type as seen in the CLAB data plotted in Figure 6.10 is associated 

with the difference in uranium mass between the two assembly types. Because BWR assemblies typically 

contain about one-half the uranium mass of PWR assemblies, they have lower decay heat values for 

similar burnups. A more consistent basis for comparing the different assemblies is the specific decay 

heat, which is decay heat divided by the initial uranium mass. This basis results in a modified distribution 

of residuals, which are compared for the GE-Morris and CLAB facilities in Figures 6.11 and 6.12. As 

observed in the earlier comparisons (see Figures 6.9 and 6.10), the PWR data from GE-Morris have 

relatively low error, while the BWR data have much larger error and exhibit a small negative bias. 

However, there are no significant correlations (i.e., trends) between the error and specific decay heat. 

Results for the CLAB assemblies are shown in Figure 6.13. The BWR residuals show little bias, being 

centered near zero. The PWR data exhibit a small positive bias relative to the BWR data, although 

this difference is minor. The bias could be due to the measurements or their subsequent analysis 

(e.g., calorimeter calibration, specific heat effect of shroud used for BWR assemblies, etc.). It could also 

be due to systematic calculational error, e.g., differences in the models or input data used to generate the 

PWR and BWR libraries for the ORIGEN-S calculations. However, the latter situation does not appear 

credible since the bias has been observed using a wide range of models, methods, and nuclear data, and 

such a bias was not observed in the GE-Morris PWR measurements involving similar assembly designs. 

The GE-Morris residual errors shown in Figure 6.11 do not show a correlation with reactor type. Both 

reactor types exhibit a negative bias; however the PWR data are much closer to zero. The BWR data 

indicate a much greater scatter and stronger negative bias, opposite to the behavior observed in the CLAB 

results. Again it is possible that such effects are due to systematic errors in either the measurements or 

calculation input (or both). 

Finally, the residual errors are examined as a function of assembly burnup, decay time, and enrichment in 

Figures 6.14–6.16. From these plots it is apparent that the errors are virtually uncorrelated with these 

variables (there are no obvious trends). This is true regardless of the measurement facility involved. Any 

slight trends or nonrandom effects are better explained by other factors (previously discussed) than by 

correlations with these quantities. 
102 

-60

-40

-20

0

20

40

0 200 400 600 800 1000

Measured Decay Heat (W)

Residual Error (W)

BWR

PWR

Figure 6.11 Error for GE-Morris measurements by reactor type 

-60

-40

-20

0

20

40

0 0.5 1 1.5 2 2.5

Measured Specific Decay Heat (kW/MTU)

Residual Error (W)

BWR

PWR

Figure 6.12 Residual error in specific decay heat for GE-Morris measurements 
103 

-10

-5

0

5

10

15

20

0 0.5 1 1.5

Measured Specific Decay Heat (kW/MTU)

Residual Error (W)

BWR

PWR

Figure 6.13 Residual error in specific decay heat for CLAB measurements 

-60

-40

-20

0

20

40

60

0 10000 20000 30000 40000 50000

Burnup (MWd/MTU)

Residual Error (W)

GE-Morris

CLAB

HEDL

Figure 6.14 Residual error correlation with assembly burnup 
104 

-60

-40

-20

0

20

40

60

0 2000 4000 6000 8000 10000

Decay Time (days)

Residual Error (W)

GE-Morris

CLAB

HEDL

Figure 6.15 Residual error correlation with assembly cooling time 

-60

-40

-20

0

20

40

60

012345

Enrichment ( wt % 235U)

Residual Error (W)

GE-Morris

CLAB

HEDL

Figure 6.16 Residual error correlation with assembly enrichment 
105 

6.2.4 Sensitivities and Uncertainties in Calculations 

Errors in code calculations may be introduced by modeling approximations and errors, numerical 

discretization, and errors in input data. Regarding modeling and numerical error, attempts have been 

made to minimize this and otherwise account for it with well-established numerical algorithms and 

thorough code QA and testing. SCALE is maintained and developed under a formal code QA plan that 

covers testing of the methods and data.58,59 The numerical method of the ORIGEN code (matrix 

exponential with Bateman solutions) has also been benchmarked against other independent numerical and 

analytical solutions for fuel irradiation and decay problems in an international code comparison60

involving the use of common nuclear reaction and decay data. A complete discussion of numerical error 

is beyond the scope of this discussion. However, errors due to the numerical methods are judged to be 

very small with respect to other sources. We therefore focus on the error in ORIGEN calculations due to 

uncertain or erroneous input data. 

A thorough evaluation of calculational error was undertaken for an earlier version of the code,61 and much 

of that study is applicable now as well. The earlier study identified a number of input parameters to 

which the decay heat was most sensitive. Foremost among them were power, irradiation time, and 

cooling time, as one would expect. Other quantities include 

1. cooling time since discharge; 

2. total irradiation time; 

3. assembly power and power history (variations over irradiation time); 

4. initial uranium mass in the assembly; 

5. initial enrichment in 235U; 

6. structural materials and other activated components; and 

7. nuclear data (cross sections, fission yields, decay constants, energy per decay Q values, energy 

per fission Q values). 

The amount of cobalt (59Co) as an impurity in assembly structural components varies widely and may 

contribute to decay heat (and gamma radiation sources) due to formation of 60Co by activation. However, 

uncertainty from this contribution is still generally small compared with other sources of error. 

The most significant quantity influencing decay heat is the assembly power, and therefore it is important 

for the code input to be as accurate as possible. In this study the cycle power was derived from the cycle 

burnup. The operating times of the reactor are very well known. Therefore, the error in the power is 

similar to the error in the cycle burnup. The final assembly burnup for U.S. fuel assemblies was obtained 

from DOE Form RW-859, which reports burnup calculated by the utilities. A limited review of the 

uncertainties in utility-declared burnup has been reported. The accuracy of reactor records was studied in 

1997 by the Electric Power Research Institute (EPRI) in its report Determination of the Accuracy of 

Utility Spent-Fuel Burnup Records.

62 The EPRI report concluded that the uncertainty associated with the 

calculated core-follow burnup values is less than 2% for Westinghouse PWR assemblies. Further 

evaluations63 using a larger sample of reactor records from nine PWR plants found uncertainties in the 

range of 2 to 5% at a 95%-confidence level. For more complex BWR assembly designs and reactor 

operations one might expect somewhat larger uncertainties for these assemblies. The relative sensitivity 

coefficients for the reactor power, or burnup (impact of changes in power on the predicted decay heat), 

are estimated in Ref. 3 to be > 1 for cooling times from 3 years to 100 years. Therefore, an error of 2% in 

the burnup value will yield an error of more than 2% in decay heat. 
106 

The total reactor power is, of course, not constant, but varies throughout its cycle history. The pattern of 

variation can be approximated by histograms representing periods of constant power. The detail 

necessary to accurately model the total power history of a given assembly has also been studied 

previously. Comparisons using early versions of the ORIGEN code indicated that using a detailed power 

history was slightly better than using individual cycle averages29,30 and that using individual cycle 

averages was much better than using a lifetime average. However, obtaining a correct detailed history 

may be problematic, and various sources for such information may yield different values, as seen in this 

work with the Dresden DN212 assembly history (see Sect. 3). The accuracy of calculations using cycleaverage power estimates (used in this work) is estimated to be accurate to better than 2% as compared to 

calculations with very detailed power simulations. 

Other important quantities include the nuclear data. Earlier sensitivity studies involving ORIGEN 

indicated only a minor importance of cross-section errors on decay heat for times less than about 

100 years. Cross sections become more importance with cooling time as the higher actinides (Pu, Am, 

and Cm) contribute in large proportions to the decay heat. The buildup of actinides is very sensitive to 

cross sections. However, the earlier studies did not address high-burnup fuels (which have higher 

actinide content) or the complex BWR assembly designs now in use. The new designs generally make 

extensive use of burnable poisons, and with axial voiding, produce higher actinide inventories than in 

previous designs. Therefore it is reasonable to assume that the effect of cross-section errors may be 

somewhat larger for modern fuels than previously estimated. 

6.3 SAS2H Analysis Results

Decay heat calculations evaluated in Appendix A were made using ORIGEN-S and assembly crosssection libraries created using 2-D TRITON models. Calculations for the Swedish assemblies were also 

performed using the 1-D SAS2H depletion analysis sequence of SCALE using the modeling procedures 

described in Sect. 5. SAS2H couples the 1-D XSDRNPM transport analysis code with the ORIGEN-S 

code to performed automated cross-section generation and fuel burnup analysis. Although the 1-D 

methods require more assembly model approximations than 2-D methods, particularly for heterogeneous 

modern BWR assembly designs, the 1-D sequence has been extensively validated and applied for a wide 

range of commercial spent fuel studies. The results reported here were obtained directly from SAS2H 

calculations; that is, a separate SAS2H calculation was performed for each assembly, rather than using the 

sequence to generate cross-section libraries that were later used in standalone ORIGEN-S calculations. 

The SAS2H results, limited in this study to the Swedish assemblies, are compared with the results of 

TRITON in Appendix C. 

The SAS2H results are similar to those obtained using the cross-section libraries generated by TRITON 

for all assemblies studied. The average difference between the SAS2H and TRITON decay heat 

predictions are less than 1% for all Swedish assemblies. The differences are also seen to be similar for 

the BWR and PWR assemblies, indicating that any modeling approximations in the 1-D SAS2H models 

for the 8 × 8, 9 × 9, SVEA assemblies do not have a large impact on the results (e.g., the results for the 

complex BWR assembly designs are similar to the less complex PWR designs). In fact, the BWR results 

are on average in closer agreement than the PWR results. However, as burnup and cooling time increase 

the differences between the results are expected to become larger as the actinides become a more 

important component of decay heat. The actinide concentrations are more sensitive to the cross section 

variations (and hence modeling approximations) than most fission products. A comparison of the SAS2H 

and TRITON results suggest some degree of correlation between the deviation and the burnup. The 

largest differences are observed for the highest burnup assemblies. 
107 

6.4 Effect of Using Assembly Average Burnup and Coolant Density

All of the results presented so far in this report were performed with simulation models that used 

assembly average data and did not account for any axial variation of burnup or coolant density. Assembly 

burnup and water moderator densities applied in the calculations were based on a volumetric average over 

the axial length of the assembly. However, decay heat is not strictly a linear function of burnup; this is 

particularly true for the decay heat component from actinides. Coolant density also influences the cross 

sections and the buildup of actinides in a nonlinear manner. To study this effect and determine the 

accuracy of models based on assembly average data, selected assemblies measured at the Swedish CLAB 

facility that had detailed axial information were simulated using an explicit multizone model. In this 

model, separate simulations of each axial zone, characterized by the zone burnup (and coolant density in 

the BWR case) were performed and the assembly decay heat then determined as the sum of the decay heat 

values obtained for each axial zone. These results were then compared to those obtained using assembly 

average parameters. 

For the PWR case only the axial variation in burnup was considered, the change in moderator density 

over the length of the assembly being relatively small. The multizone model for the PWR assemblies 

considered 24 axial regions of equal volume for which burnup information was available. In the BWR 

case both axial burnup and coolant density variations were available for 25 regions. PWR and BWR 

assemblies were selected to include representative assembly configurations and burnup values. The axial 

burnup and coolant density values, obtained from Ref. 39, are based on utility records. The specific 

assemblies and their main characteristics are listed in Table 6.2. 

Table 6.2 Assemblies considered for axial decay heat studies 

Reactor 

type 

Reactor 

name 

Assembly 

design 

Assembly 

ID 

Enrichment 

(wt% 235U) 

Burnup 

(GWd/MTU) 

 Oskarshamn 2 8 × 8 1389 2.2 19.481 

BWR Barsebäck 1 8 × 8 9329 2.9 41.094 

 Oskarshamn 2 SVEA-64 12684 2.9 46.648 

 Oskarshamn 3 SVEA-100 13628 2.7 35.619 

 Ringhals 2 15 × 15 G23 3.2 35.633 

PWR Ringhals 2 15 × 15 F32 3.2 50.962 

 Ringhals 3 17 × 17 5A3 2.1 19.699 

 Ringhals 3 17 × 17 1C5 3.1 38.484 

The ratio of the decay heat value obtained for the one-zone model to the decay heat for the multizone 

model was calculated for cooling times from discharge to 150 years. This ratio, Rz, is illstrated in 

Figure 6.17 for the PWR and the BWR assemblies. Rz is observed by be less than 1 for the PWR 

assemblies selected in this study for cooling times up to about 40 to 60 years, depending on the burnup 

and other characteristics of the assembly. For longer times Rz is greater than 1 (i.e., the one-zone model 

overpredicts relative to the explicit multizone model). However, the effect is seen to be relatively small; 

less than 1% over all times studied. The magnitude of the effect depends on the cooling time and burnup 

and is most pronounced for shorter cooling times and increases with burnup. 

The effect of the axial simulations for the BWR assemblies is similar to that for the PWR assemblies, 

however, the magnitude is larger. At longer cooling times of interest to interim storage the use of average 

assembly burnup and coolant density leads to an overestimation of the decay heat by up to 2.5%. 
108 

Although detailed axial calculations may not be practical for routine decay heat analysis, and this effect is 

relatively small (the magnitude in many cases being smaller than the experimental uncertainties), the user 

needs to be aware of the bias introduced in assembly models that use average data instead of detailed 

axial data. Importantly, it can be seen from Figure 6.17 that there is no single average value for the 

assembly burnup (or coolant density) that will yield results that are equivalent to a detailed axial 

calculation. 

Figure 6.17 Ratio of decay heat calculated using assembly average parameters 

to that obtained using explicit axial calculations for selected Swedish BWR and 

PWR assemblies 

0 20 40 60 80 100 120 140 160

0.97

0.98

0.99

1.00

1.01

1.02

0 20 40 60 80 100 120 140 160

0.97

0.98

0.99

1.00

1.01

1.02

Decay heat ratio (one-zone / multi-zone)

 1389

 9329

 12684

 13826

BW R

 5A3

 1C5

 G23

 F32

D ecay tim e (years)

PW R
109 

7. SUMMARY AND CONCLUSIONS 

The fuel burnup and decay code ORIGEN-S of the SCALE 5 system has been validated for decay heat 

predictions involving full-length spent fuel assemblies using calorimeter measurements performed at 

GE-Morris operation, the HEDL facility, and Swedish CLAB interim spent fuel storage facility. In all, 

more than 190 assembly measurements have been evaluated. The most recent CLAB measurements 

greatly extended the range of data available for code validation beyond that previously available in the 

United States. A statistical analysis of the results has been carried out based on the evaluation of all 

measurements, not just the most recent measurements. The decision to include all available experimental 

data resulted in a larger code uncertainty estimate than would have been obtained by using only the 

CLAB data, which have significantly lower experimental errors than the earlier GE-Morris and HEDL 

measurements. The use of all measurement results establishes the widest validated range of application 

for the code. 

In addition to the full-length assembly measurements, the ORIGEN-S code was benchmarked against 

measurements of decay heat following the fission of the actinides 233U, 235U, 238U, 239Pu, 241Pu, and 232Th 

at short cooling times of importance to accident analyses. The results indicate that the calculations are 

generally within the range of experimental uncertainty for the measurements evaluated. However, several 

earlier calorimetric measurements not used in this study yield systematically higher decay heat values as 

compared with the spectroscopic measurements considered here. Further quantitative analysis of these 

measurements was beyond the scope of this work. 

The remainder of this section summarizes the main conclusions that have been drawn from the study of 

the spent fuel assembly experiments. 

• The validation covers a wide range of assembly designs, initial enrichments extending up to 

4.0 wt % 235U, burnup values up to 50,962 MWd/MTU, and cooling times extending to 27 years 

after discharge. 

• The results show that the decay heat predicted by ORIGEN-S is generally within the range of 

uncertainty of the measurements. An evaluation of the calorimeter uncertainty based on 

repeatability of measurements gives measurement uncertainty values that are very facility 

dependent: ± 17 W for GE-Morris, ± 95 W for HEDL, and ± 4 W for CLAB. The approximate 

relative errors are: ± 5%, 10%, and 2%, respectively. 

• The deviations between calculations and measurements (residuals) are observed to be normally 

distributed with a mean error (bias) of 0.7 W, a value close to zero. The variance of the residuals 

is seen to be approximately independent of the decay heat rate. The standard deviation assigned 

to the calculation based on all residuals is ± 17.6 W, a value very similar to the accuracy of 

GE-Morris measurement data, but higher than that of the CLAB data. A statistical analysis 

indicates that the relative errors obtained using the measured data from all facilities combined are 

not normally distributed. Therefore caution is needed when representing the calculation 

uncertainty in relative units. 

• The ability to validate the accuracy of the code predictions is limited by the accuracy of the 

measurements; calculational uncertainty can not be demonstrated to be lower than the uncertainty 

inherent in the data itself. However, based on the calorimeter uncertainties estimated from repeat 

measurements, it is likely that the errors in the measurements themselves represent a significant 

component of the differences between calculations and measurements. 
110 

• The present level of uncertainty assigned to the calculations is not likely to be further reduced 

significantly by performing additional measurements if the results of the GE-Morris and HEDL 

measurements are retained because of the relatively large deviations associated with some of 

these data. 

• The uncertainty of ± 17.6 W is relatively small for decay heat rates of 200 W or more 

(e.g., < 10%). However, at lower decay heat rates the error becomes more significant and may 

become unacceptably large at longer cooling times. Again, it is not possible with the current data 

to determine absolutely whether this error is associated with the accuracy of the measurements or 

with the calculations. 

• Based on an analysis of the data, there is no compelling rational to evaluate the bias and 

uncertainty of the PWR assembly results separately from the BWR results. Any correlation 

between the two types of assemblies would be difficult to establish with any confidence given the 

variability of the measurement data. 

• Simulation of assembly decay heat as a point model (one zone with axial and radial parameters 

averaged) leads to a small underprediction in the decay heat of typically < 2%, caused primarily 

by the averaging of the axial burnup profile and void profile (for BWRs) of the assembly. 

• Very similar calculated results were obtained using cross-section libraries generated by 1-D 

(SAS2H) and 2-D (TRITON) physics models of the fuel assembly lattice. In general, the 

difference due to this change in the physics models was found to be less than 2%. 

Based on these findings the following recommendations are made: 

• The ability to further reduce the current level of calculation uncertainty is limited by the 

availability of measurement data with lower errors. The use of the GE-Morris and HEDL data in 

the present study clearly leads to much larger uncertainties because of the large scatter in many of 

the measurements. At such time as the CLAB measurement range can be extended to cover the 

full range of the previous U.S. measurements, a reanalysis of uncertainties should be conducted 

using only the more precise CLAB data. Based on the results for the existing CLAB 

measurements, replacing the GE-Morris and HEDL data could potentially reduce the present 

level of code uncertainty from ± 17.6 W to approximately ± 10 W or possibly lower. 

• Acquisition of additional measurement data for CLAB assemblies that cover low decay heat 

values of the earlier GE-Morris data, and higher decay heat range represented by the HEDL data, 

are desirable to improve and reduce the uncertainty in decay heat predictions. 
111 

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Las Vegas NV, 2004. 

A-1 

APPENDIX A

SPENT FUEL ASSEMBLY BENCHMARK DATA 

SUMMARY OF EXPERIMENTAL AND CALCULATED RESULTS
A-2 
A-3 

Table A.1 Summary of measured and predicted PWR fuel assembly decay heat (U.S. measurement

Reactor Assembly type Assembly ID Enrichment (wt % 235U) Uranium (kg) Final burnup (MWd/t) Discharge date Measurement date Cooling time (days) Measured (W) Calculate

(W) 

Point Beach 14×14 C-52 3.397 386.54 31,914 3/3/1977 8/23/1981 1635 724.0 718.Point Beach 14×14 C-56 3.397 386.80 38,917 3/3/1977 8/24/1981 1634 921.0 917.Point Beach 14×14 C-64 3.397 386.63 39,384 3/3/1977 8/23/1981 1633 931.0 931.Point Beach 14×14 C-66 3.397 386.54 35,433 3/3/1977 8/20/1981 1630 846.0 834.Point Beach 14×14 C-67 3.397 386.45 38,946 3/3/1977 8/19/1981 1629 934.0 919.Point Beach 14×14 C-68 3.397 386.36 37,059 3/3/1977 8/20/1981 1630 874.0 876.San Onofre 14×14s*** C-01 3.865 361.72 26,540 6/2/1973 8/30/1981 3011 359.0 360.San Onofre 14×14s C-16** 3.865 361.72 28,462 6/2/1973 8/31/1981 3012 384.0 388.San Onofre 14×14s C-19 3.865 361.72 30,426 6/2/1973 8/30/1981 3011 418.0 417.San Onofre 14×14s C-20 3.865 361.72 32,363 6/2/1973 8/29/1981 3011 456.0 447.San Onofre 14×14s D-01 4.005 363.64 31,393 3/14/1975 8/27/1981 2358 499.0 489.San Onofre 14×14s D-46 4.005 363.64 32,318 3/14/1975 8/29/1981 2360 510.0 507.San Onofre 14×14s E-18 4.005 363.98 32,357 9/30/1976 8/28/1981 1794 635.0 630.San Onofre 14×14s F-04 3.996 372.32 30,429 9/15/1978 8/28/1981 1078 934.0 927.Turkey Point 15×15 B-43 2.559 447.79 25,595 10/26/1975 9/11/1980 1782 637.0 619.Turkey Point 15×15 D-15 2.557 456.12 28,430 11/24/1977 7/8/1980 957 1423.0 1444.Turkey Point 15×15 D-15* 2.557 456.12 28,430 11/24/1977 1/6/1981 1139 1126.0 1175.Turkey Point 15×15 D-15* 2.557 456.12 28,430 11/24/1977 7/28/1983 2072 625.0 623.Turkey Point 15×15 D-22 2.557 458.00 26,485 11/24/1977 7/9/1980 958 1284.0 1280.Turkey Point 15×15 D-34 2.557 455.24 27,863 11/24/1977 4/1/1980 859 1550.0 1590.

* Indicates repeat measurements on the same assembly. 

 ** Measured value from static mode operation (recirculation mode value rejected). *** “s” indicates stainless steel clad fuel. 
A-4 

Table A.2 Summary of measured and predicted BWR fuel assembly decay heat (U.S. measurement

Reactor Assembly 

type 

Assembly 

ID 

Enrichment

(wt % 

235U) 

Uranium

(kg) 

Final 

burnup 

(MWd/t) 

Discharge 

date 

Measurement 

date 

Cooling 

time 

(days) 

Measured 

(W) 

Calculated 

(W) 

Cooper GE 7×7 CZ102 1.090 195.48 11,667 9/17/1977 9/25/1984 2565 62.3 80.0 Cooper GE 7×7 CZ102* 1.090 195.48 11,667 9/17/1977 12/14/1984 2645 70.4 78.7 Cooper GE 7×7 CZ147 2.500 190.31 26,709 4/20/1981 11/4/1984 1294 276.7 282.7 Cooper GE 7×7 CZ148 2.500 190.22 26,310 4/20/1981 10/23/1984 1282 273.5 280.2 Cooper GE 7×7 CZ182 2.500 190.09 26,823 5/21/1982 9/27/1984 860 342.6 344.1 Cooper GE 7×7 CZ195 2.500 190.68 26,391 4/20/1981 10/29/1984 1288 255.5 277.2 Cooper GE 7×7 CZ205 2.500 190.72 25,344 5/21/1982 9/24/1984 857 324.0 314.0 Cooper GE 7×7 CZ205* 2.500 190.72 25,344 5/21/1982 10/4/1984 867 361.5 311.2 Cooper GE 7×7 CZ205* 2.500 190.72 25,344 5/21/1982 10/8/1984 871 343.5 310.0 Cooper GE 7×7 CZ205* 2.500 190.72 25,344 5/21/1982 10/9/1984 872 353.2 309.8 Cooper GE 7×7 CZ205* 2.500 190.72 25,344 5/21/1982 10/23/1984 886 331.8 306.0 Cooper GE 7×7 CZ205* 2.500 190.72 25,344 5/21/1982 10/24/1984 887 338.7 305.7 Cooper GE 7×7 CZ205* 2.500 190.72 25,344 5/21/1982 10/29/1984 892 327.5 304.4 Cooper GE 7×7 CZ205* 2.500 190.72 25,344 5/21/1982 11/2/1984 896 313.1 303.3 Cooper GE 7×7 CZ205* 2.500 190.72 25,344 5/21/1982 11/5/1984 899 311.4 302.5 Cooper GE 7×7 CZ205* 2.500 190.72 25,344 5/21/1982 12/6/1984 930 314.0 294.7 Cooper GE 7×7 CZ205* 2.500 190.72 25,344 5/21/1982 12/12/1984 936 331.2 293.2 Cooper GE 7×7 CZ205* 2.500 190.72 25,344 5/21/1982 12/22/1984 946 317.1 290.9 Cooper GE 7×7 CZ205* 2.500 190.72 25,344 5/21/1982 5/13/1985 1088 289.7 261.8 Cooper GE 7×7 CZ205* 2.500 190.72 25,344 5/21/1982 5/28/1985 1103 308.0 259.1 Cooper GE 7×7 CZ209 2.500 190.38 25,383 5/21/1982 10/28/1984 891 279.5 277.2 Cooper GE 7×7 CZ211 2.500 190.82 26,679 4/20/1981 10/1/1984 1260 296.0 289.3 Cooper GE 7×7 CZ211* 2.500 190.82 26,679 4/20/1981 5/20/1985 1491 240.3 252.5 Cooper GE 7×7 CZ222 2.500 190.90 26,692 5/21/1982 11/4/1984 898 355.7 327.5 Cooper GE 7×7 CZ225 2.500 190.51 25,796 5/21/1982 10/2/1984 865 333.5 304.2 Cooper GE 7×7 CZ239 2.500 189.57 27,246 5/21/1982 10/30/1984 893 366.5 337.3 

* Indicates repeat measurements on the same assembly. 
A-5 

Table A.2 Summary of measured and predicted BWR fuel assembly decay heat (U.S. measurements only

Reactor Assembly 

type 

Assembly 

ID 

Enrichment

(wt % 235U)

Uranium

(kg) 

Final burnup

(MWd/t) 

Discharge 

date 

Measurement 

date 

Cooling time

(days) 

Measured

(W) 

Calculated

(W) 

Cooper GE 7×7 CZ246 2.500 189.81 27,362 5/21/1982 11/2/1984 896 320.9 341.8 Cooper GE 7×7 CZ246* 2.500 189.81 27,362 5/21/1982 11/5/1984 899 341.7 340.9 Cooper GE 7×7 CZ259 2.500 190.20 26,466 4/20/1981 10/29/1984 1288 247.6 277.1 Cooper GE 7×7 CZ259* 2.500 190.20 26,466 4/20/1981 12/20/1984 1340 288.5 268.4 Cooper GE 7×7 CZ259* 2.500 190.20 26,466 4/20/1981 5/14/1985 1485 254.1 247.8 Cooper GE 7×7 CZ264 2.500 190.89 26,496 4/20/1981 10/23/1984 1282 263.8 280.3 Cooper GE 7×7 CZ277 2.500 189.49 26,747 4/20/1981 10/27/1984 1286 262.7 276.8 Cooper GE 7×7 CZ277* 2.500 189.49 26,747 4/20/1981 5/26/1985 1497 243.0 245.7 Cooper GE 7×7 CZ286 2.500 189.95 27,141 5/21/1982 12/6/1984 930 278.4 308.3 Cooper GE 7×7 CZ286* 2.500 189.95 27,141 5/21/1982 5/28/1985 1103 284.2 273.1 Cooper GE 7×7 CZ296 2.500 190.50 26,388 4/20/1981 11/3/1984 1293 256.7 283.8 Cooper GE 7×7 CZ296* 2.500 190.50 26,388 4/20/1981 5/21/1985 1492 251.9 252.2 Cooper GE 7×7 CZ302 2.500 190.00 26,594 4/20/1981 10/24/1984 1283 285.6 276.7 Cooper GE 7×7 CZ308 2.500 189.78 25,815 5/21/1982 11/1/1984 895 269.7 282.2 Cooper GE 7×7 CZ311 2.500 189.91 27,392 5/21/1982 10/27/1984 890 356.9 320.3 Cooper GE 7×7 CZ315 2.500 189.96 26,881 5/21/1982 12/8/1984 932 328.0 299.3 Cooper GE 7×7 CZ318 2.500 189.32 26,568 5/21/1982 12/7/1984 931 277.6 280.8 Cooper GE 7×7 CZ331 2.500 190.36 21,332 3/31/1978 9/24/1984 2369 162.8 156.3 Cooper GE 7×7 CZ331* 2.500 190.36 21,332 3/31/1978 12/21/1984 2457 180.1 153.0 Cooper GE 7×7 CZ337 2.500 189.90 26,720 5/21/1982 11/1/1984 895 347.7 325.9 Cooper GE 7×7 CZ337* 2.500 189.90 26,720 5/21/1982 5/24/1985 1099 300.4 278.4 Cooper GE 7×7 CZ342 2.500 190.16 27,066 5/21/1982 12/7/1984 931 280.1 302.1 Cooper GE 7×7 CZ342* 2.500 190.16 27,066 5/21/1982 5/26/1985 1101 300.0 268.6 Cooper GE 7×7 CZ346 2.500 190.23 28,048 5/21/1982 10/27/1984 890 388.7 354.0 Cooper GE 7×7 CZ348 2.500 190.38 27,480 5/21/1982 10/31/1984 894 342.8 335.0 Cooper GE 7×7 CZ351 2.500 190.02 25,753 5/21/1982 12/10/1984 934 313.8 281.4 

* Indicates repeat measurements on the same assembly. 
A-6 

Table A.2 Summary of measured and predicted BWR fuel assembly decay heat (U.S. measurements only

Reactor Assembly 

type 

Assembly 

ID 

Enrichment

(wt % 235U)

Uranium

(kg) 

Final burnup

(MWd/t) 

Discharge 

date 

Measurement 

date 

Cooling time

(days) 

Measured

(W) 

Calculated

(W) 

Cooper GE 7×7 CZ355 2.500 190.60 25,419 5/21/1982 10/28/1984 891 290.5 278.4 Cooper GE 7×7 CZ357 2.500 190.19 27,140 5/21/1982 12/8/1984 932 320.3 308.1 Cooper GE 7×7 CZ369 2.500 190.20 26,575 5/21/1982 10/25/1984 888 347.6 323.4 Cooper GE 7×7 CZ370 2.500 190.23 26,342 4/20/1981 9/28/1984 1257 293.6 278.6 Cooper GE 7×7 CZ372 2.500 190.01 25,848 4/20/1981 9/27/1984 1256 294.3 263.7 Cooper GE 7×7 CZ379 2.500 190.18 25,925 5/21/1982 11/4/1984 898 287.4 281.3 Cooper GE 7×7 CZ398 2.500 189.83 27,478 5/21/1982 10/27/1984 890 372.0 338.8 Cooper GE 7×7 CZ415 2.500 189.72 25,863 4/20/1981 9/26/1984 1255 289.3 272.8 Cooper GE 7×7 CZ416 2.500 189.43 27,460 5/21/1982 10/31/1984 894 319.8 318.0 Cooper GE 7×7 CZ429 2.500 190.07 27,641 5/21/1982 10/26/1984 889 385.6 348.3 Cooper GE 7×7 CZ430 2.500 189.93 26,824 5/21/1982 10/31/1984 894 353.3 324.1 Cooper GE 7×7 CZ433 2.500 190.02 25,977 4/20/1981 9/26/1984 1255 287.4 264.1 Cooper GE 7×7 CZ433* 2.500 190.02 25,977 4/20/1981 5/21/1985 1492 256.7 231.3 Cooper GE 7×7 CZ460 2.500 190.18 26,511 5/21/1982 12/9/1984 933 313.5 292.4 Cooper GE 7×7 CZ466 2.500 189.86 26,077 5/21/1982 9/28/1984 861 302.1 292.5 Cooper GE 7×7 CZ468 2.500 189.78 26,757 5/21/1982 12/11/1984 935 325.3 300.2 Cooper GE 7×7 CZ472 2.500 190.12 25,957 5/21/1982 9/26/1984 859 325.0 319.7 Cooper GE 7×7 CZ473 2.500 189.76 26,517 5/21/1982 12/10/1984 934 293.2 281.2 Cooper GE 7×7 CZ498 2.500 189.69 26,482 5/21/1982 10/25/1984 888 359.4 324.3 Cooper GE 7×7 CZ508 2.500 190.68 26,357 5/21/1982 12/9/1984 933 310.0 293.4 Cooper GE 7×7 CZ515 2.500 190.48 25,737 4/20/1981 9/25/1984 1254 294.0 272.4 Cooper GE 7×7 CZ515* 2.500 190.48 25,737 4/20/1981 10/26/1984 1285 296.0 266.9 Cooper GE 7×7 CZ526 2.500 190.54 27,596 5/21/1982 10/1/1984 864 395.4 356.7 Cooper GE 7×7 CZ526* 2.500 190.54 27,596 5/21/1982 5/22/1985 1097 321.8 296.0 Cooper GE 7×7 CZ528 2.500 190.81 25,714 4/20/1981 10/25/1984 1284 297.6 267.3 Cooper GE 7×7 CZ531 2.500 189.90 26,699 5/21/1982 10/30/1984 893 347.2 323.2 

* Indicates repeat measurements on the same assembly. 
A-7 

Table A.2 Summary of measured and predicted BWR fuel assembly decay heat (U.S. measurements only

Reactor Assembly 

type 

Assembly 

ID 

Enrichment

(wt % 235U)

Uranium

(kg) 

Final burnup

(MWd/t) 

Discharge 

date 

Measurement 

date 

Cooling time

(days) 

Measured

(W) 

Calculated

(W) 

Cooper GE 7×7 CZ536 2.500 190.17 26,589 4/20/1981 9/27/1984 1256 295.2 262.2 Cooper GE 7×7 CZ542 2.500 189.99 26,691 5/21/1982 12/8/1984 932 311.9 295.2 Cooper GE 7×7 CZ545 2.500 190.47 26,668 5/21/1982 12/11/1984 935 295.2 285.0 Dresden GE 7×7 DN212 2.130 194.70 5,280 9/17/1976 10/3/1984 2938 31.2 34.8 Dresden GE 7×7 DN212* 2.130 194.70 5,280 9/17/1976 10/18/1984 2953 19.5 34.7 Monticello GE 7×7 MT116 2.250 193.53 18,039 9/12/1975 6/10/1985 3559 114.9 109.2 Monticello GE 7×7 MT123 2.250 193.53 13,027 3/15/1974 6/5/1985 4100 66.8 75.7 Monticello GE 7×7 MT123* 2.250 193.53 13,027 3/15/1974 6/8/1985 4103 95.3 75.7 Monticello GE 7×7 MT123* 2.250 193.53 13,027 3/15/1974 6/11/1985 4106 65.9 75.7 Monticello GE 7×7 MT133 2.250 193.53 20,994 9/12/1975 5/29/1985 3547 152.6 129.1 Monticello GE 7×7 MT133* 2.250 193.53 20,994 9/12/1975 6/6/1985 3555 129.0 129.0 Monticello GE 7×7 MT133* 2.250 193.53 20,994 9/12/1975 6/9/1985 3558 154.8 128.9 Monticello GE 7×7 MT133* 2.250 193.53 20,994 9/12/1975 6/12/1985 3561 106.7 128.9 Monticello GE 7×7 MT190 2.250 193.53 15,143 9/12/1975 6/8/1985 3557 99.2 92.0 Monticello GE 7×7 MT228 2.250 193.53 12,123 9/12/1975 5/30/1985 3548 101.0 73.3 Monticello GE 7×7 MT228* 2.250 193.53 12,123 9/12/1975 6/7/1985 3556 71.2 73.3 Monticello GE 7×7 MT228* 2.250 193.53 12,123 9/12/1975 6/11/1985 3560 76.4 73.3 Monticello GE 7×7 MT264 2.250 193.53 9,160 9/12/1975 6/5/1985 3554 46.1 54.4 

* Indicates repeat measurements on the same assembly. 
A-8 

Table A.3 Summary of measured and predicted PWR fuel assembly decay heat (SKB measurement

Reactor Assembly 

type 

Assembly 

ID 

Enrichment 

(wt % 235U)

Uranium

(kg) 

Final burnup 

(MWd/t) 

Discharge 

date 

Measurement 

date 

Cooling time

(days) 

Measured

(W) 

Calculate

(W) 

Ringhals 2 15×15 C01 3.095 455.79 36,688 4/4/1981 6/10/2004 8468 415.8 423.6 Ringhals 2 15×15 C12 3.095 453.74 36,385 4/4/1981 4/6/2004 8403 410.3 418.9 Ringhals 2 15×15 C20 3.095 454.76 35,720 4/4/1985 4/14/2004 6950 415.8 433.9 Ringhals 2 15×15 C20* 3.095 454.76 35,720 4/4/1985 4/15/2004 6951 426.0 433.8 Ringhals 2 15×15 C20* 3.095 454.76 35,720 4/4/1985 4/16/2004 6952 428.9 433.8 Ringhals 2 15×15 C20* 3.095 454.76 35,720 4/4/1985 4/23/2004 6959 435.7 433.7 Ringhals 2 15×15 D27 3.252 432.59 39,676 4/28/1983 4/26/2004 7669 456.1 455.6 Ringhals 2 15×15 D38 3.252 434.21 39,403 5/6/1982 4/5/2004 8005 442.3 446.1 Ringhals 2 15×15 E38 3.199 433.59 33,973 5/6/1982 3/30/2004 7999 376.3 376.7 Ringhals 2 15×15 E38* 3.199 433.59 33,973 5/6/1982 3/31/2004 8000 374.3 376.7 Ringhals 2 15×15 E40 3.199 434.24 34,339 5/6/1982 6/14/2004 8075 381.3 380.6 Ringhals 2 15×15 F14 3.197 436.38 34,009 4/28/1983 6/18/2004 7722 381.8 385.6 Ringhals 2 15×15 F21 3.197 435.94 36,273 4/13/1984 6/23/2004 7376 420.9 420.5 Ringhals 2 15×15 F25 3.197 437.29 35,352 4/28/1983 6/21/2004 7725 396.7 403.5 Ringhals 2 15×15 F32 3.197 436.99 50,962 5/12/1988 5/28/2004 5860 692.0 697.8 Ringhals 2 15×15 G11 3.188 436.18 35,463 4/4/1985 5/24/2004 6990 416.4 415.7 Ringhals 2 15×15 G23 3.206 436.13 35,633 4/4/1985 5/18/2004 6984 420.6 422.2 Ringhals 2 15×15 I09 3.203 437.35 40,188 5/12/1988 5/17/2004 5849 507.9 518.7 Ringhals 2 15×15 I24 3.203 429.60 34,294 4/30/1986 5/26/2004 6601 410.1 405.9 Ringhals 2 15×15 I25 3.203 433.06 36,859 4/25/1987 4/13/2004 6198 445.8 454.0 

* Indicates repeat measurements on the same assembly 
A-9 

Table A.3 Summary of measured and predicted PWR fuel assembly decay heat (SKB measurements only

Reactor Assembly 

type 

Assembly 

ID 

Enrichment 

(wt % 235U)

Uranium

(kg) 

Final burnup 

(MWd/t) 

Discharge 

date 

Measurement 

date 

Cooling time

(days) 

Measured

(W) 

Calculate

(W) 

Ringhals 3 17×17 0C9 3.101 457.64 38,442 5/30/1986 5/6/2004 6551 491.2 508.3 Ringhals 3 17×17 0E2 3.103 463.60 41,628 7/7/1988 6/16/2004 5823 587.9 588.6 Ringhals 3 17×17 0E6 3.103 461.77 35,993 7/7/1988 6/22/2004 5829 487.8 491.8 Ringhals 3 17×17 1C2 3.101 459.05 33,318 5/30/1986 5/14/2004 6559 417.7 426.2 Ringhals 3 17×17 1C5 3.101 457.99 38,484 5/30/1986 6/17/2004 6593 499.2 508.3 Ringhals 3 17×17 1E5 3.103 463.90 34,638 7/7/1988 6/11/2004 5818 468.8 472.4 Ringhals 3 17×17 2A5 2.100 462.03 20,107 5/11/1984 5/3/2004 7297 233.8 240.2 Ringhals 3 17×17 2C2 3.101 459.49 36,577 5/30/1986 5/5/2004 6550 466.5 476.2 Ringhals 3 17×17 3C1 3.101 458.43 36,572 5/30/1986 4/30/2004 6545 470.2 475.1 Ringhals 3 17×17 3C5 3.101 458.87 38,373 5/30/1986 4/28/2004 6543 501.4 508.8 Ringhals 3 17×17 3C9 3.101 459.14 36,560 5/30/1986 5/7/2004 6552 468.4 475.5 Ringhals 3 17×17 4C4 3.101 459.05 33,333 5/30/1986 5/27/2004 6572 422.0 426.3 Ringhals 3 17×17 4C7 3.101 458.26 38,370 5/30/1986 5/4/2004 6549 498.7 507.9 Ringhals 3 17×17 5A3 2.100 461.48 19,699 5/11/1984 6/13/2003 6972 237.7 238.5 Ringhals 3 17×17 5A3* 2.100 461.48 19,699 5/11/1984 6/16/2003 6975 236.7 238.5 Ringhals 3 17×17 5A3* 2.100 461.48 19,699 5/11/1984 6/18/2003 6977 243.4 238.5 Ringhals 3 17×17 5A3* 2.100 461.48 19,699 5/11/1984 4/27/2004 7291 230.9 234.9 Ringhals 3 17×17 5A3* 2.100 461.48 19,699 5/11/1984 5/10/2004 7304 230.3 234.8 

* Indicates repeat measurements on the same assembly 
A-10 

Table A.4 Summary of measured and predicted BWR fuel assembly decay heat (SKB measurement

Reactor Assembly 

type 

Assembly 

ID 

Enrichment

(wt % 235U)

Uranium

(kg) 

Final burnup

(MWd/t) 

Discharge 

date 

Measurement 

date 

Cooling 

time 

(days) 

Measured

(W) 

Calculate

(W) 

Barsebäck 1 8×8−1 9329 2.920 178.77 41,127 9/17/1988 6/2/2003 5371 222.8 221.2 Barsebäck 1 8×8−1 9329* 2.920 178.77 41,127 9/17/1988 6/4/2003 5373 224.4 221.1 Barsebäck 1 8×8−1 9329* 2.920 178.77 41,127 9/17/1988 11/20/2003 5542 218.7 218.6 Barsebäck 1 8×8−1 10288 2.950 179.16 35,218 9/17/1988 11/12/2003 5534 185.8 183.5 Barsebäck 2 8×8−1 14076 3.150 179.57 40,010 7/2/1992 12/9/2003 4177 240.3 235.2 Forsmark 1 8×8−1 3838 2.086 177.90 25,669 7/10/1992 12/10/2003 4170 126.8 131.7 Forsmark 1 8×8−1 3838* 2.086 177.90 25,669 7/10/1992 12/11/2003 4171 125.9 131.6 Forsmark 1 8×8−2 KU0100 2.976 174.92 34,193 8/17/1990 1/9/2004 4893 185.3 177.9 Forsmark 1 9×9−5 KU0269 2.938 177.02 35,113 8/17/1990 1/19/2004 4903 192.7 192.2 Forsmark 1 9×9−5 KU0278 2.939 177.13 35,323 5/24/1991 12/22/2003 4595 195.4 194.4 Forsmark 1 9×9−5 KU0282 2.939 177.10 37,896 5/24/1991 12/1/2003 4574 218.5 212.2 Forsmark 2 8×8−1 5535 2.095 177.69 19,944 7/15/1988 12/18/2003 5634 84.6 92.4 Forsmark 2 SVEA 64 11494 2.920 181.09 32,431 7/15/1988 12/2/2003 5618 166.0 165.8 Forsmark 2 SVEA 64 11495 2.910 181.07 32,431 7/15/1988 11/7/2003 5593 167.6 166.6 Forsmark 2 SVEA 64 13775 2.850 181.34 32,837 7/12/1991 12/19/2003 4543 178.4 178.8 Forsmark 3 SVEA 100 13847 2.769 180.67 31,275 7/13/1990 11/13/2003 4871 170.3 168.4 Forsmark 3 SVEA 100 13847* 2.769 180.67 31,275 7/13/1990 11/13/2003 4871 169.6 168.4 Forsmark 3 SVEA 100 13848 2.769 180.67 31,275 7/13/1990 11/24/2003 4882 170.7 168.3 Ringhals 1 8×8−1 1177 2.642 180.59 36,242 8/2/1985 11/26/2003 6690 177.9 173.6 Ringhals 1 8×8−1 1186 2.640 180.52 30,498 8/2/1985 11/10/2003 6674 140.8 142.2 Ringhals 1 8×8−1 6423 2.900 177.70 35,109 8/5/1988 2/12/2004 5669 174.2 175.3 Ringhals 1 8×8−1 6432 2.894 177.52 36,861 8/5/1988 6/10/2003 5422 185.5 186.5 Ringhals 1 8×8−1 6432* 2.894 177.52 36,861 8/5/1988 6/12/2003 5424 189.5 186.5 

* Indicates repeat measurements on the same assembly. 
A-11 

Table A.4 Summary of measured and predicted BWR fuel assembly decay heat (SKB measurements only

Reactor Assembly 

type 

Assembly 

ID 

Enrichment

(wt % 

235U) 

Uranium

(kg) 

Final 

burnup 

(MWd/t) 

Discharge 

date 

Measurement 

date 

Cooling 

time 

(days) 

Measured

(W) 

Calculate

(W) 

Ringhals 1 8×8−1 6432* 2.894 177.52 36,861 8/5/1988 2/13/2004 5670 184.4 183.6 Ringhals 1 8×8−1 6432* 2.894 177.52 36,861 8/5/1988 2/23/2004 5680 182.8 183.5 Ringhals 1 8×8−1 6432* 2.894 177.52 36,861 8/5/1988 2/24/2004 5681 185.2 183.5 Ringhals 1 8×8−1 6432* 2.894 177.52 36,861 8/5/1988 3/1/2004 5687 181.6 183.4 Ringhals 1 8×8−1 6432* 2.894 177.52 36,861 8/5/1988 3/2/2004 5688 182.0 183.4 Ringhals 1 8×8−1 6454 2.898 177.68 37,236 8/15/1986 2/17/2004 6395 186.3 180.6 Ringhals 1 8×8−1 8327 2.904 177.54 37,851 8/6/1991 3/10/2004 4600 196.9 200.7 Ringhals 1 8×8−1 8331 2.910 177.69 35,903 9/15/1989 3/11/2004 5291 187.0 183.3 Ringhals 1 8×8−1 8332 2.895 177.52 34,977 8/5/1988 3/4/2004 5690 168.1 172.1 Ringhals 1 8×8−1 8338 2.911 177.60 34,830 8/5/1988 3/9/2004 5695 169.5 172.0 Oskarshamn 2 8×8−1 1377 2.201 183.58 14,546 5/13/1977 1/22/2004 9750 56.2 59.2 Oskarshamn 2 8×8−1 1389 2.201 183.65 19,481 7/15/1981 11/28/2003 8171 83.9 83.2 Oskarshamn 2 8×8−1 1546 2.201 183.97 24,470 8/19/1983 1/16/2004 7455 108.1 110.6 Oskarshamn 2 8×8−1 1696 2.201 184.25 20,870 8/19/1983 12/3/2003 7411 92.4 93.6 Oskarshamn 2 8×8−1 1704 2.201 184.02 19,437 7/23/1982 12/8/2003 7808 84.0 84.0 Oskarshamn 2 8×8−1 2995 2.699 179.38 29,978 7/15/1981 1/7/2004 8211 130.5 131.2 Oskarshamn 2 8×8−1 6350 2.875 179.00 27,675 6/7/1985 12/4/2003 6754 126.9 128.9 Oskarshamn 2 8×8−1 6350* 2.875 179.00 27,675 6/7/1985 12/5/2003 6755 129.4 128.9 Oskarshamn 2 SVEA 64 12684 2.902 182.32 46,648 8/2/1991 12/16/2003 4519 282.7 279.1 Oskarshamn 3 8×8−1 12078 2.577 177.36 25,160 7/8/1988 11/18/2003 5611 120.2 125.6 Oskarshamn 3 

SVEA 

100 13628 2.711 180.77 35,619 6/24/1991 

1/8/2004 

4581 194.0 196.2 

Oskarshamn 3 

SVEA 

100 13630 2.711 180.78 40,363 6/24/1991 

12/12/2003 

4554 235.7 232.3 

* Indicates repeat measurements on the same assembly. 

B-1 

APPENDIX B

ARPLIB INPUT FILE EXAMPLE 

TRITON INPUT FILES FOR SELECTED ASSEMBLIES
B-2 
B-3 

ARPLIB input file for San Onofre PWR 14 × 14 library 

In this example two libraries (named w14_e35.lib and w14_e40.lib), representing two discrete 

enrichments (3.5 and 4.0 w t%), are thinned from 25 burnup-dependent cross section sets down to 11 final 

cross section sets and saved with a library name suffix arplib. The #shell commands are used to 

copy the libraries between the home area ($RTNDIR) and the SCALE working directory ($TMPDIR) on 

a Unix or Linux operating system. On a Windows OS the cp command is replace with copy, and the 

home directory environment variable $RTNDIR becomes %RTNDIR%. 

#shell 

cp $RTNDIR/w14*.lib ./ 

end 

=arplib 

-1 

25 

1 1 1 1 1 1 1 0 0 0 0 1 0 0 0 0 1 0 0 0 1 0 0 0 1 

2 

w14_e35.lib 

w14_e35.arplib 

w14_e40.lib 

w14_e40.arplib 

end 

#shell 

cp w14*.arplib $RTNDIR 

end 
B-4 

TRITON input file for PWR 14 × 14 San Onofre assembly 

=t-depl parm=nitawl 

PWR 14x14 for San Onofre assembly, 4.0 wt% U-235 

44groupndf5 

'======== 

read comp 

'fuel 

uo2 1 0.93 810 92234 0.03473 

 92235 4.00 

 92236 0.01840 

 92238 95.9469 end 

' important nuclides from Table 1 of ORNL/TM-12294/V1 

zr-94 1 0 1.00e-20 810 end 

tc-99 1 0 1.00e-20 810 end 

ru-106 1 0 1.00e-20 810 end 

rh-103 1 0 1.00e-20 810 end 

rh-105 1 0 1.00e-20 810 end 

xe-131 1 0 1.00e-20 810 end 

cs-133 1 0 1.00e-20 810 end 

cs-134 1 0 1.00e-20 810 end 

ce-144 1 0 1.00e-20 810 end 

pr-143 1 0 1.00e-20 810 end 

nd-143 1 0 1.00e-20 810 end 

nd-145 1 0 1.00e-20 810 end 

nd-147 1 0 1.00e-20 810 end 

pm-147 1 0 1.00e-20 810 end 

sm-149 1 0 1.00e-20 810 end 

sm-151 1 0 1.00e-20 810 end 

sm-152 1 0 1.00e-20 810 end 

eu-153 1 0 1.00e-20 810 end 

eu-154 1 0 1.00e-20 810 end 

eu-155 1 0 1.00e-20 810 end 

cm-245 1 0 1.00e-20 810 end 

cm-246 1 0 1.00e-20 810 end 

cm-247 1 0 1.00e-20 810 end 

cm-248 1 0 1.00e-20 810 end 

bk-249 1 0 1.00e-20 810 end 

cf-249 1 0 1.00e-20 810 end 

cf-250 1 0 1.00e-20 810 end 

cf-251 1 0 1.00e-20 810 end 

cf-252 1 0 1.00e-20 810 end 

u-232 1 0 1.00e-20 810 end 

u-233 1 0 1.00e-20 810 end 

u-237 1 0 1.00e-20 810 end 

pu-236 1 0 1.00e-20 810 end 

pu-237 1 0 1.00e-20 810 end 

pu-238 1 0 1.00e-20 810 end 

pu-239 1 0 1.00e-20 810 end 

pu-240 1 0 1.00e-20 810 end 

pu-241 1 0 1.00e-20 810 end 

pu-242 1 0 1.00e-20 810 end 

pu-243 1 0 1.00e-20 810 end 

pu-244 1 0 1.00e-20 810 end 

'clad 

ss304 2 1 615 end 

cobalt 2 0 1.00e-20 615 end 

'water moderator with 500 ppm B 

h2o 3 den=0.7179 1 576.5 end 

arbmb 0.7179 1 1 0 0 5000 100 3 500e-06 576.5 end 

'gap 

n 4 den=0.00125 1 615 end 

'guide tube 

ss304 5 1 576.5 end 

cobalt 5 0 1.00e-20 576.5 end 

end comp 

'======== 

read celldata 

 latticecell squarepitch pitch=1.4122 3 

 fueld=0.9741 1 

 gapd=0.9881 4 

 cladd=1.0719 2 

 end 

end celldata 

'======== 

read depletion 

 -1 2 5 

end depletion 

'======== 
B-5 

read burndata 

 power=40.0 burn=1e-15 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

end burndata 

'======== 

read model 

Westinghouse 14x14 

'-------- 

read parm 

 fillmix=3 epseigen=1e-4 epsinner=-1e-4 

 epsouter=1e-4 cmfd=yes xycmfd=4 echo=yes 

end parm 

'-------- 

read materials 

 1 1 ! fuel ! end 

 2 1 ! clad ! end 

 3 2 ! water ! end 

 4 0 ! gap ! end 

 5 1 ! guide tube ! end 

end materials 

'-------- 

read geom 

unit 1 

com='fuel rod' 

 cylinder 10 .48705 

 cylinder 20 .49405 

 cylinder 30 .53595 

 cuboid 40 4p0.7061 

 media 1 1 10 

 media 4 1 20 -10 

 media 2 1 30 -20 

 media 3 1 40 -30 

 boundary 40 4 4 

unit 2 

com='guide tube' 

 cylinder 10 .64895 

 cylinder 20 .67945 

 cuboid 40 4p0.7061 

 media 3 1 10 

 media 5 1 20 -10 

 media 3 1 40 -20 

 boundary 40 4 4 

global unit 10 

 cuboid 10 19.94 0.0 19.94 0.0 

 array 1 10 place 1 1 0.7907 0.7907 

 media 3 1 10 

 boundary 10 56 56 

end geom 

'-------- 

read array 

 ara=1 nux=14 nuy=14 typ=cuboidal 

 fill 

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 

 1 1 2 1 1 2 1 1 2 1 1 2 1 1 

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 

 1 1 1 1 2 1 1 1 1 2 1 1 1 1 

 1 1 2 1 1 1 1 1 1 1 1 2 1 1 
B-6 

 1 1 1 1 1 1 1 2 1 1 1 1 1 1 

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 

 1 1 2 1 1 1 1 1 1 1 1 2 1 1 

 1 1 1 1 2 1 1 1 1 2 1 1 1 1 

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 

 1 1 2 1 1 2 1 1 2 1 1 2 1 1 

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 

 end fill 

end array 

'-------- 

read bounds 

 all=refl 

end bounds 

'-------- 

end model 

'======== 

end 
B-7 

TRITON input file for BWR 7 × 7 Cooper assembly* 

=t-depl parm=nitawl 

BWR 7x7, 3.0 wt% U-235, 0.5 gg/cc coolant density 

44groupndf5 

'======== 

read comp 

'fuel 1 cell type (no gadolinia) 

uo2 1 den=9.73 1 840 92234 0.02543 

 92235 3.00 

 92236 0.01380 

 92238 96.9608 end 

'clad 

zirc4 2 1 620 end 

'water coolant 

h2o 3 den=0.5 1 558 end 

'fuel 2 cell type (with gadolinia 3&) 

uo2 5 den=9.73 0.97 840 92234 0.02543 

 92235 3.00 

 92236 0.01380 

 92238 96.9608 end 

arbmgd 9.73 2 0 1 0 64000 2 8016 3 5 0.03 840 end 

'clad 

zirc2 6 1 620 end 

'water coolant 

h2o 7 den=0.5 1 558 end 

'channel tube 

zirc4 9 1 558 end 

'water in water rod and channel moderator 

h2o 10 den=0.7627 1 552 end 

'fuel 3 cell type (with gadolinia 4%) 

uo2 11 den=9.73 0.96 840 92234 0.02543 

 92235 3.00 

 92236 0.01380 

 92238 96.9608 end 

arbmgd 9.73 2 0 1 0 64000 2 8016 3 11 0.04 840 end 

'clad 

zirc2 12 1 620 end 

'water coolant 

h2o 13 den=0.5 1 558 end 

end comp 

'======== 

read celldata 

 latticecell squarepitch pitch=1.875 3 

 fueld=1.242 1 

 cladd=1.430 2 end 

 latticecell squarepitch pitch=1.875 7 

 fueld=1.242 5 

 cladd=1.430 6 end 

 latticecell squarepitch pitch=1.875 13 

 fueld=1.242 11 

 cladd=1.430 12 end 

end celldata 

'======== 

read depletion 

 1 5 11 

end depletion 

'======== 

read burndata 

 power=40.0 burn=1e-15 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 
B-8 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

end burndata 

'======== 

read model 

BWR 7x7 

'------- 

read parm 

 fillmix=10 epseigen=1e-4 epsinner=-1e-4 

 epsouter=1e-4 cmfd=yes xycmfd=4 echo=yes 

end parm 

'------- 

read materials 

 1 1 ! fuel ! end 

 2 1 ! clad ! end 

 3 2 ! coolant ! end 

 5 1 ! fuel-Gd (3%) ! end 

 9 1 ! channel tube ! end 

10 2 ! moderator ! end 

11 1 ! fuel-Gd (4%) ! end 

end materials 

'------- 

read geom 

unit 1 

com='fuel pin' 

 cylinder 10 .621 

 cylinder 20 .715 

 cuboid 40 4p0.9375 

 media 1 1 10 

 media 2 1 20 -10 

 media 3 1 40 -20 

 boundary 40 4 4 

unit 2 

com='fuel wih Gd 3%' 

 cylinder 10 .621 

 cylinder 20 .715 

 cuboid 40 4p0.9375 

 media 5 1 10 

 media 2 1 20 -10 

 media 3 1 40 -20 

 boundary 40 4 4 

unit 3 

com='fuel wih Gd 4%' 

 cylinder 10 .621 

 cylinder 20 .715 

 cuboid 40 4p0.9375 

 media 11 1 10 

 media 2 1 20 -10 

 media 3 1 40 -20 

 boundary 40 4 4 

global unit 5 

com='assembly' 

 cuboid 51 14.1825 1.0575 14.1825 1.0575 

 cuboid 52 14.3825 0.8575 14.3825 0.8575 

 cuboid 53 15.24 0 15.24 0 

 array 1 51 place 1 1 1.995 1.995 

 media 3 1 51 

 media 9 1 52 -51 

 media 10 1 53 -52 

 boundary 53 36 36 

end geom 

'------- 

read array 

 ara=1 nux=7 nuy=7 

 fill 

 1 1 1 1 1 1 1 

 1 1 2 1 1 3 1 

 1 1 1 1 1 1 1 

 1 1 1 2 1 1 1 

 1 1 1 1 1 2 1 

 1 3 1 1 1 1 1 

 1 1 1 1 1 1 1 

 end fill 

end array 

'------- 

read bounds 

 all=refl 

end bounds 
B-9 

'------- 

end data 

'======== 

end 

*

 Trace nuclides need to be included in fuel composition, as for the 7 × 7 assembly; not shown here for brevity. 
B-10 

TRITON input file for PWR 17 × 17 Ringhals-3 assembly* 

=t-depl parm=nitawl 

PWR 17x17 Type 1, Swedish data, 3.0 wt% U-235 

44groupndf5 

'======== 

read comp 

'fuel 

uo2 1 den=10.27 1 900 92234 0.02543 

 92235 3.00 

 92236 0.01380 

 92238 96.9608 end 

'clad 

zirc4 2 1 600 end 

'water moderator with 650 ppm B 

h2o 3 den=0.72 1 577 end 

arbmb 0.72 1 1 0 0 5000 100 3 650e-06 577 end 

'gap 

n 4 den=0.00125 1 600 end 

'guide tube 

zirc4 5 1 577 end 

end comp 

'======== 

read celldata 

 latticecell squarepitch pitch=1.26 3 

 fueld=0.8191 1 

 gapd=0.8358 4 

 cladd=0.950 2 

 end 

end celldata 

'======== 

read depletion 

 1 

end depletion 

'======== 

read burndata 

 power=40.0 burn=1e-15 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

end burndata 

'======== 

read model 

PWR 17x17 

'------- 

read parm 

 fillmix=3 epseigen=1e-4 epsinner=-1e-4 

 epsouter=1e-4 run=yes echo=yes 

end parm 

'------- 

read materials 

 1 1 ! fuel ! end 

 2 1 ! clad ! end 

 3 2 ! water ! end 

 4 0 ! gap ! end 

 5 1 ! guide tube ! end 

end materials 

'------- 
B-11 

read geom 

unit 1 

com='fuel rod' 

 cylinder 10 .40955 

 cylinder 20 .41790 

 cylinder 30 .4750 

 cuboid 40 4p.63 

 media 1 1 10 

 media 4 1 20 -10 

 media 2 1 30 -20 

 media 3 1 40 -30 

 boundary 40 4 4 

unit 5 

com='guide tube' 

 cylinder 10 .56135 

 cylinder 20 .602 

 cuboid 40 4p.63 

 media 3 1 10 

 media 5 1 20 -10 

 media 3 1 40 -20 

 boundary 40 4 4 

unit 11 

com='right half of fuel rod' 

 cylinder 10 .40955 chord +x=0 

 cylinder 20 .41790 chord +x=0 

 cylinder 30 .4750 chord +x=0 

 cuboid 40 .63 0.0 2p.63 

 media 1 1 10 

 media 4 1 20 -10 

 media 2 1 30 -20 

 media 3 1 40 -30 

 boundary 40 2 4 

unit 12 

com='top half of fuel rod' 

 cylinder 10 .40955 chord +y=0 

 cylinder 20 .41790 chord +y=0 

 cylinder 30 .4750 chord +y=0 

 cuboid 40 2p.63 .63 0.0 

 media 1 1 10 

 media 4 1 20 -10 

 media 2 1 30 -20 

 media 3 1 40 -30 

 boundary 40 4 2 

unit 51 

com='right half of guide tube' 

 cylinder 10 .56135 chord +x=0 

 cylinder 20 .602 chord +x=0 

 cuboid 40 .63 0.0 2p.63 

 media 3 1 10 

 media 5 1 20 -10 

 media 3 1 40 -20 

 boundary 40 2 4 

unit 52 

com='top half of guide tube' 

 cylinder 10 .56135 chord +y=0 

 cylinder 20 .602 chord +y=0 

 cuboid 40 2p.63 .63 0.0 

 media 3 1 10 

 media 5 1 20 -10 

 media 3 1 40 -20 

 boundary 40 4 2 

unit 53 

com='1/4 instrument tube' 

 cylinder 10 .56135 chord +x=0 chord +y=0 

 cylinder 20 .602 chord +x=0 chord +y=0 

 cuboid 40 .63 0.0 .63 0.0 

 media 3 1 10 

 media 5 1 20 -10 

 media 3 1 40 -20 

 boundary 40 2 2 

global unit 10 

 cuboid 10 10.75 0.0 10.75 0.0 

 array 1 10 place 1 1 0 0 

 media 3 1 10 

 boundary 10 34 34 

end geom 

'------- 

read array 

 ara=1 nux=9 nuy=9 typ=cuboidal 

 fill 

 53 12 12 52 12 12 52 12 12 
B-12 

 11 1 1 1 1 1 1 1 1 

 11 1 1 1 1 1 1 1 1 

 51 1 1 5 1 1 5 1 1 

 11 1 1 1 1 1 1 1 1 

 11 1 1 1 1 5 1 1 1 

 51 1 1 5 1 1 1 1 1 

 11 1 1 1 1 1 1 1 1 

 11 1 1 1 1 1 1 1 1 

 end fill 

end array 

'------- 

read bounds 

 all=refl 

end bounds 

'------- 

end model 

'======== 

end 

*

 Trace nuclides need to be included in fuel composition, as for the 7 × 7 assembly; not shown here for brevity. 
B-13 

TRITON input file for PWR 15 × 15 Ringhals-3 assembly* 

=t-depl parm=nitawl 

PWR 15x15 Type 2, Swedish data, 3.0 wt% U-235 

44groupndf5 

'========== 

read comp 

'fuel 

uo2 1 den=10.1373 1 900 92234 0.02543 

 92235 3.00 

 92236 0.01380 

 92238 96.9608 end 

'clad 

zirc4 2 1 600 end 

'water moderator with 650 ppm B 

h2o 3 den=0.72 1 577 end 

arbmb 0.72 1 1 0 0 5000 100 3 650e-06 577 end 

'gap 

n 4 den=0.00125 1 600 end 

'guide tube 

zirc4 5 1 577 end 

end comp 

'======== 

read celldata 

 latticecell squarepitch pitch=1.430 3 

 fueld=0.911 1 

 gapd=0.930 4 

 cladd=1.075 2 

 end 

end celldata 

'======== 

read depletion 

 1 

end depletion 

'======== 

read burndata 

 power=40.0 burn=1e-15 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

end burndata 

'======== 

read model 

PWR 15x15 

'-------- 

read parm 

 fillmix=3 epseigen=1e-4 epsinner=-1e-4 epsouter=1e-4 

 cmfd=yes xycmfd=4 echo=yes 

end parm 

'-------- 

read materials 

 1 1 ! fuel ! end 

 2 1 ! clad ! end 

 3 2 ! water ! end 

 4 0 ! gap ! end 

 5 1 ! guide tube ! end 

end materials 

'-------- 
B-14 

read geom 

unit 1 

com='fuel pin' 

 cylinder 10 .4555 

 cylinder 20 .4650 

 cylinder 30 .5375 

 cuboid 40 4p0.715 

 media 1 1 10 

 media 4 1 20 -10 

 media 2 1 30 -20 

 media 3 1 40 -30 

 boundary 40 4 4 

unit 5 

com='guide tube' 

 cylinder 10 .6515 

 cylinder 20 .6945 

 cuboid 40 4p0.715 

 media 3 1 10 

 media 5 1 20 -10 

 media 3 1 40 -20 

 boundary 40 4 4 

unit 11 

com='right half of fuel pin' 

 cylinder 10 .4555 chord +x=0 

 cylinder 20 .4650 chord +x=0 

 cylinder 30 .5375 chord +x=0 

 cuboid 40 0.715 0.0 2p0.715 

 media 1 1 10 

 media 4 1 20 -10 

 media 2 1 30 -20 

 media 3 1 40 -30 

 boundary 40 2 4 

unit 12 

com='top half of fuel pin' 

 cylinder 10 .4555 chord +y=0 

 cylinder 20 .4650 chord +y=0 

 cylinder 30 .5375 chord +y=0 

 cuboid 40 2p0.715 0.715 0.0 

 media 1 1 10 

 media 4 1 20 -10 

 media 2 1 30 -20 

 media 3 1 40 -30 

 boundary 40 4 2 

unit 51 

com='right half of guide tube' 

 cylinder 10 .6515 chord +x=0 

 cylinder 20 .6945 chord +x=0 

 cuboid 40 0.715 0.0 2p0.715 

 media 3 1 10 

 media 5 1 20 -10 

 media 3 1 40 -20 

 boundary 40 2 4 

unit 52 

com='top half of guide tube' 

 cylinder 10 .6515 chord +y=0 

 cylinder 20 .6945 chord +y=0 

 cuboid 40 2p0.715 0.715 0.0 

 media 3 1 10 

 media 5 1 20 -10 

 media 3 1 40 -20 

 boundary 40 4 2 

unit 53 

com='1/4 instrument tube' 

 cylinder 10 .6515 chord +x=0 chord +y=0 

 cylinder 20 .6945 chord +x=0 chord +y=0 

 cuboid 40 0.715 0.0 0.715 0.0 

 media 3 1 10 

 media 5 1 20 -10 

 media 3 1 40 -20 

 boundary 40 2 2 

global unit 10 

 cuboid 10 10.75 0.0 10.75 0.0 

 array 1 10 place 1 1 0 0 

 media 3 1 10 

 boundary 10 30 30 

end geom 

'-------- 

read array 

 ara=1 nux=8 nuy=8 typ=cuboidal 

 fill 

 53 12 12 12 52 12 12 12 
B-15 

 11 1 1 1 1 1 1 1 

 11 1 1 1 1 5 1 1 

 11 1 1 5 1 1 1 1 

 51 1 1 1 1 1 1 1 

 11 1 5 1 1 5 1 1 

 11 1 1 1 1 1 1 1 

 11 1 1 1 1 1 1 1 

 end fill 

end array 

'-------- 

read bounds 

 all=refl 

end bounds 

'-------- 

end model 

'========== 

end 

*

 Trace nuclides need to be included in fuel composition, as for the 7x7 assembly; not shown here for brevity. 
B-16 

TRITON input file for BWR 8 × 8 Ringhals-1 assembly* 

=t-depl parm=nitawl 

BWR 8x8 based Type 2 with 4 Gd rods, 3.0 wt% U-235, 0.5 g/cc coolant density 

44groupndf5 

'======== 

read comp 

'fuel 1 cell (no gadolinia) 

uo2 1 den=10.31619 1 900 92235 3.00 

 92238 97.0 end 

'clad 

zirc2 2 1 573 end 

'gap 

n 3 den=1.25e-3 573 end 

'coolant 

h2o 4 den=0.5 1 552 end 

'fuel 2 cell (with gadolinia) 

uo2 5 den=10.31619 0.98 900 92235 3.00 

 92238 97.0 end 

arbmgd 10.31619 2 0 1 0 64000 2 8016 3 5 0.02 900 end 

'clad 

zirc2 6 1 573 end 

'gap 

n 7 den=1.25e-3 573 end 

'coolant 

h2o 8 den=0.5 1 552 end 

'water rod tube 

zirc2 9 1 552 end 

'water in water rod 

h2o 10 den=0.75 1 552 end 

'channel box 

zirc4 11 1 552 end 

'channel moderator 

h2o 12 den=0.75 1 552 end 

end comp 

'======== 

read celldata 

latticecell squarepitch pitch=1.63 4 

 fueld=1.044 1 

 cladd=1.225 2 

 gapd=1.065 3 end 

latticecell squarepitch pitch=1.63 8 

 fueld=1.044 5 

 cladd=1.225 6 

 gapd=1.065 7 end 

end celldata 

'======== 

read depletion 

 1 5 

end depletion 

'======== 

read burndata 

 power=40.0 burn=1e-15 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

end burndata 

'======== 
B-17 

read model 

BWR 8x8 

'-------- 

read parm 

 fillmix=12 epseigen=1e-4 epsinner=-1e-4 

 epsouter=1e-4 cmfd=yes xycmfd=4 echo=yes 

end parm 

'-------- 

read materials 

 1 1 ! fuel ! end 

 2 1 ! clad ! end 

 3 0 ! gap ! end 

 4 2 ! coolant ! end 

 5 1 ! fuel-Gd ! end 

 9 1 ! water rod tube ! end 

10 2 ! water in water rod ! end 

11 1 ! channel box ! end 

12 1 ! channel moderator ! end 

end materials 

'-------- 

read geom 

unit 1 

com='fuel pin' 

 cylinder 10 0.522 

 cylinder 20 0.5325 

 cylinder 30 0.6125 

 cuboid 40 4p0.815 

 media 1 1 10 

 media 3 1 20 -10 

 media 2 1 30 -20 

 media 4 1 40 -30 

 boundary 40 4 4 

unit 2 

com='fuel wih Gd pin' 

 cylinder 10 0.522 

 cylinder 20 0.5325 

 cylinder 30 0.6125 

 cuboid 40 4p0.815 

 media 5 1 10 

 media 3 1 20 -10 

 media 2 1 30 -20 

 media 4 1 40 -30 

 boundary 40 4 4 

unit 3 

com='water rod' 

 cylinder 10 0.5325 

 cylinder 20 0.6125 

 cuboid 30 4p0.815 

 media 10 1 10 

 media 9 1 20 -10 

 media 4 1 30 -20 

 boundary 30 4 4 

unit 11 

com='fuel pin - NW corner' 

 cylinder 10 0.497 origin x=0.075 y=-0.075 

 cylinder 20 0.5075 origin x=0.075 y=-0.075 

 cylinder 30 0.5875 origin x=0.075 y=-0.075 

 cuboid 40 4p0.815 

 media 1 1 10 

 media 3 1 20 -10 

 media 2 1 30 -20 

 media 4 1 40 -30 

 boundary 40 4 4 

unit 12 

com='fuel pin - NNW and WNW off-corner' 

 cylinder 10 0.497 origin x=0.025 y=-0.025 

 cylinder 20 0.5075 origin x=0.025 y=-0.025 

 cylinder 30 0.5875 origin x=0.025 y=-0.025 

 cuboid 40 4p0.815 

 media 1 1 10 

 media 3 1 20 -10 

 media 2 1 30 -20 

 media 4 1 40 -30 

 boundary 40 4 4 

unit 13 

com='fuel pin - NE corner' 

 cylinder 10 0.497 origin x=-0.075 y=-0.075 

 cylinder 20 0.5075 origin x=-0.075 y=-0.075 

 cylinder 30 0.5875 origin x=-0.075 y=-0.075 

 cuboid 40 4p0.815 

 media 1 1 10 
B-18 

 media 3 1 20 -10 

 media 2 1 30 -20 

 media 4 1 40 -30 

 boundary 40 4 4 

unit 14 

com='fuel pin - NNE and ENE off-corner' 

 cylinder 10 0.497 origin x=-0.025 y=-0.025 

 cylinder 20 0.5075 origin x=-0.025 y=-0.025 

 cylinder 30 0.5875 origin x=-0.025 y=-0.025 

 cuboid 40 4p0.815 

 media 1 1 10 

 media 3 1 20 -10 

 media 2 1 30 -20 

 media 4 1 40 -30 

 boundary 40 4 4 

unit 15 

com='fuel pin - SE corner' 

 cylinder 10 0.497 origin x=-0.075 y=0.075 

 cylinder 20 0.5075 origin x=-0.075 y=0.075 

 cylinder 30 0.5875 origin x=-0.075 y=0.075 

 cuboid 40 4p0.815 

 media 1 1 10 

 media 3 1 20 -10 

 media 2 1 30 -20 

 media 4 1 40 -30 

 boundary 40 4 4 

unit 16 

com='fuel pin - SSE and ESE off-corner' 

 cylinder 10 0.497 origin x=-0.025 y=0.025 

 cylinder 20 0.5075 origin x=-0.025 y=0.025 

 cylinder 30 0.5875 origin x=-0.025 y=0.025 

 cuboid 40 4p0.815 

 media 1 1 10 

 media 3 1 20 -10 

 media 2 1 30 -20 

 media 4 1 40 -30 

 boundary 40 4 4 

unit 17 

com='fuel pin - SW corner' 

 cylinder 10 0.497 origin x=0.075 y=0.075 

 cylinder 20 0.5075 origin x=0.075 y=0.075 

 cylinder 30 0.5875 origin x=0.075 y=0.075 

 cuboid 40 4p0.815 

 media 1 1 10 

 media 3 1 20 -10 

 media 2 1 30 -20 

 media 4 1 40 -30 

 boundary 40 4 4 

unit 18 

com='fuel pin - SSW and WSW off-corner' 

 cylinder 10 0.497 origin x=0.025 y=0.025 

 cylinder 20 0.5075 origin x=0.025 y=0.025 

 cylinder 30 0.5875 origin x=0.025 y=0.025 

 cuboid 40 4p0.815 

 media 1 1 10 

 media 3 1 20 -10 

 media 2 1 30 -20 

 media 4 1 40 -30 

 boundary 40 4 4 

global unit 20 

com='assembly' 

 cuboid 51 4p6.72 

 cuboid 52 4p6.95 

 cuboid 53 4p7.665 

 array 1 51 place 1 1 -5.705 -5.705 

 media 4 1 51 

 media 11 1 52 -51 

 media 12 1 53 -52 

 boundary 53 40 40 

end geom 

'-------- 

read array 

 ara=1 nux=8 nuy=8 

 fill 

17 18 1 1 1 1 16 15 

18 1 1 1 1 1 1 16 

1 1 2 1 1 2 1 1 

1 1 1 1 1 1 1 1 

1 1 1 1 3 1 1 1 

1 1 2 1 1 2 1 1 

12 1 1 1 1 1 1 14 
B-19 

11 12 1 1 1 1 14 13 

 end fill 

end array 

'-------- 

read bounds 

 all=refl 

end bounds 

'-------- 

end data 

'======== 

end 

*

 Trace nuclides need to be included in fuel composition, as for the 7x7 assembly; not shown here for brevity. 
B-20 

TRITON input file for BWR 9 × 9 Forsmark-1 assembly* 

=t-depl parm=nitawl 

Forsmark-1 BWR 9x9-5 assembly, 2.938 wt% U-235, 0.5 g/cc coolant density 

44groupndf5 

'======= 

read comp 

'fuel cell 

uo2 1 den=10.1444 1 900 92234 0.02486 

 92235 2.938 

 92236 0.01351 

 92238 97.02363 end 

'clad 

zirc2 2 1 573 end 

' gap 

n 3 den=1.25e-3 573 end 

'coolant 

h2o 4 den=0.5 1 552 end 

' fuel 2 cell type (with gadolinia) 

uo2 5 den=10.1444 0.975 900 92234 0.02486 

 92235 2.938 

 92236 0.01351 

 92238 97.02363 end 

arbmgd 10.1444 2 0 1 0 64000 2 8016 3 5 0.025 900 end 

'clad 

zirc2 6 1 573 end 

'gap 

n 7 den=1.25e-3 573 end 

'coolant 

h2o 8 den=0.5 1 552 end 

'water rod tube 

zirc2 9 1 552 end 

'water in water rod 

h2o 10 den=0.74 1 552 end 

'channel box 

zirc4 11 1 552 end 

'channel moderator 

h2o 12 den=0.669 1 552 end 

b-10 12 0 7.15-6 552 end 

end comp 

'======= 

read celldata 

latticecell squarepitch pitch=1.445 4 

 fueld=0.95 1 

 cladd=1.1 2 

 gapd=0.967 3 end 

latticecell squarepitch pitch=1.445 8 

 fueld=0.95 5 

 cladd=1.1 6 

 gapd=0.967 7 end 

end celldata 

'======= 

read depletion 

 1 5 

end depletion 

'======= 

read burndata 

 power=40.0 burn=1e-15 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 
B-21 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

end burndata 

'======= 

read model 

Forsmark-1 BWR 9x9-5 

'-------- 

read parm 

 fillmix=12 epseigen=1e-4 epsinner=-1e-4 

 epsouter=1e-4 cmfd=yes xycmfd=4 echo=yes 

end parm 

'-------- 

read materials 

 1 1 ! fuel ! end 

 2 1 ! clad ! end 

 3 0 ! gap ! end 

 4 2 ! coolant ! end 

 5 1 ! fuel-Gd ! end 

 9 1 ! water rod tube ! end 

10 2 ! water in water rod ! end 

11 1 ! channel box ! end 

12 1 ! channel moderator ! end 

end materials 

'-------- 

read geom 

unit 1 

com='fuel pin' 

 cylinder 10 0.475 

 cylinder 20 0.4835 

 cylinder 30 0.55 

 cuboid 40 4p0.7225 

 media 1 1 10 

 media 3 1 20 -10 

 media 2 1 30 -20 

 media 4 1 40 -30 

 boundary 40 4 4 

unit 2 

com='fuel pin (Gd)' 

 cylinder 10 0.475 

 cylinder 20 0.4835 

 cylinder 30 0.55 

 cuboid 40 4p0.7225 

 media 5 1 10 

 media 3 1 20 -10 

 media 2 1 30 -20 

 media 4 1 40 -30 

 boundary 40 4 4 

unit 3 

com='water rod' 

 cylinder 10 0.67 

 cylinder 20 0.75 

 cuboid 30 4p0.7225 

 media 10 1 10 

 media 9 1 20 -10 

 media 4 1 30 -20 

 boundary 30 4 4 

global unit 20 

com='assembly' 

 cuboid 51 4p6.72 

 cuboid 52 4p6.95 

 cuboid 53 4p7.62 

 array 1 51 place 5 5 0. 0. 

 media 4 1 51 

 media 11 1 52 -51 

 media 12 1 53 -52 

 boundary 53 40 40 

end geom 

'-------- 

read array 

 ara=1 nux=9 nuy=9 

 fill 

 1 1 1 1 1 1 1 1 1 

 1 2 1 1 1 1 2 1 1 

 1 1 1 3 1 1 1 2 1 

 1 1 1 1 3 1 1 1 1 

 1 1 1 1 3 3 1 1 1 

 1 1 1 1 1 1 3 1 1 

 1 1 2 1 1 1 1 1 1 

 1 2 1 1 1 1 1 2 1 

 1 1 1 1 1 1 1 1 1 
B-22 

 end fill 

end array 

'-------- 

read bounds 

 all=refl 

end bounds 

'-------- 

end data 

'======= 

end 

*

 Trace nuclides need to be included in fuel composition, as for the 7x7 assembly; not shown here for brevity. 
B-23 

TRITON input file for BWR SVEA-64 assembly* 

=t-depl parm=nitawl 

BWR SVEA 64, 3.0 wt% U-235, 0.5 g/cc coolant density 

44groupndf5 

'======== 

read comp 

'fuel 1 cell type (no gadolinia) 

uo2 1 den=10.36 1 900 92234 0.02543 

 92235 3.00 

 92236 0.01380 

 92238 96.9608 end 

'clad 

zirc2 2 1 573 end 

'water coolant 

h2o 3 den=0.5 1 552 end 

'gap 

he-4 4 den=0.82e-3 1 573 end 

'fuel 2 cell type (with gadolinia 3.5%) 

uo2 5 den=10.36 0.9745 900 92234 0.02543 

 92235 3.00 

 92236 0.01380 

 92238 96.9608 end 

arbmgd 10.36 2 0 1 0 64000 2 8016 3 5 0.0255 900 end 

'clad 

zirc2 6 1 573 end 

'water coolant 

h2o 7 den=0.5 1 552 end 

'gap 

he-4 8 den=0.82e-3 1 573 end 

'channel box 

zirc4 9 1 552 end 

'water moderator 

h2o 10 den=0.75 1 552 end 

end comp 

'======== 

read celldata 

 latticecell squarepitch pitch=1.58 3 

 fueld=1.044 1 

 cladd=1.225 2 

 gapd=1.065 4 end 

 latticecell squarepitch pitch=1.58 7 

 fueld=1.044 5 

 cladd=1.225 6 

 gapd=1.065 8 end 

end celldata 

'======== 

read depletion 

 1 5 

end depletion 

'======== 

read burndata 

 power=40.0 burn=1e-15 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

end burndata 

'======== 
B-24 

read model 

BWR SVEA-64 

'-------- 

read parm 

 fillmix=10 epseigen=1e-4 epsinner=-1e-4 

 epsouter=1e-4 cmfd=yes xycmfd=4 echo=yes 

end parm 

'-------- 

read materials 

 1 1 ! fuel ! end 

 2 1 ! clad ! end 

 3 2 ! coolant ! end 

 4 0 ! gap ! end 

 5 1 ! fuel-Gd (2.55%) ! end 

 9 1 ! channel box ! end 

10 2 ! moderator ! end 

end materials 

'-------- 

read geom 

unit 1 

com='fuel pin' 

 cylinder 10 .522 

 cylinder 20 .5325 

 cylinder 30 .6125 

 cuboid 40 4p0.79 

 media 1 1 10 

 media 4 1 20 -10 

 media 2 1 30 -20 

 media 3 1 40 -30 

 boundary 40 4 4 

unit 2 

com='fuel wih Gd 2.55%' 

 cylinder 10 .522 

 cylinder 20 .5325 

 cylinder 30 .6125 

 cuboid 40 4p0.79 

 media 5 1 10 

 media 4 1 20 -10 

 media 2 1 30 -20 

 media 3 1 40 -30 

 boundary 40 4 4 

unit 3 

com='spacer rod' 

 cylinder 20 .5325 

 cylinder 30 .6125 

 cuboid 40 4p0.79 

 media 10 1 20 

 media 2 1 30 -20 

 media 3 1 40 -30 

 boundary 40 4 4 

unit 4 

com='sub-assembly NE' 

 cuboid 41 4p3.295 

 cuboid 42 3.295 -3.375 3.295 -3.375 

 array 1 41 place 2 2 -0.79 -0.79 

 media 3 1 41 

 media 9 1 42 -41 

 boundary 42 16 16 

unit 5 

com='sub-assembly NW' 

 cuboid 51 4p3.295 

 cuboid 52 3.375 -3.295 3.295 -3.375 

 array 2 51 place 2 2 -0.79 -0.79 

 media 3 1 51 

 media 9 1 52 -51 

 boundary 52 16 16 

unit 6 

com='sub-assembly SW' 

 cuboid 61 4p3.295 

 cuboid 62 3.375 -3.295 3.375 -3.295 

 array 3 61 place 2 2 -0.79 -0.79 

 media 3 1 61 

 media 9 1 62 -61 

 boundary 62 16 16 

unit 7 

com='sub-assembly SE' 

 cuboid 71 4p3.295 

 cuboid 72 3.295 -3.375 3.375 -3.295 

 array 4 71 place 2 2 -0.79 -0.79 

 media 3 1 71 

 media 9 1 72 -71 
B-25 

 boundary 72 16 16 

global unit 8 

com='assembly' 

 cuboid 81 4p6.87 

 cuboid 82 4p6.98 

 cuboid 83 4p7.7 

 hole 4 origin x=3.575 y=3.575 

 hole 5 origin x=-3.575 y=3.575 

 hole 6 origin x=-3.575 y=-3.575 

 hole 7 origin x=3.575 y=-3.575 

 media 10 1 81 

 media 9 1 82 -81 

 media 10 1 83 -82 

 boundary 83 24 24 

end geom 

'-------- 

read array 

 ara=1 nux=4 nuy=4 

 fill 

 1 1 1 1 

 1 2 1 1 

 1 1 1 1 

 1 1 1 1 

 end fill 

 ara=2 nux=4 nuy=4 

 fill 

 1 1 1 1 

 1 1 1 1 

 1 2 1 1 

 1 1 1 1 

 end fill 

 ara=3 nux=4 nuy=4 

 fill 

 1 1 1 1 

 1 1 1 1 

 1 1 2 1 

 1 1 1 1 

 end fill 

 ara=4 nux=4 nuy=4 

 fill 

 1 1 1 1 

 1 1 2 1 

 1 3 1 1 

 1 1 1 1 

 end fill 

end array 

'-------- 

read bounds 

 all=refl 

end bounds 

end data 

'======== 

end 

*

 Trace nuclides need to be included in fuel composition, as for the 7 × 7 assembly; not shown here for brevity. 
B-26 

TRITON input file for BWR SVEA-100 assembly* 

=t-depl parm=nitawl 

BWR SVEA-100, 3.0 wt% U-235, 0.5 g/cc coolant density 

44groupndf5 

'======== 

read comp 

'fuel 1 cell type (no gadolinia) 

uo2 1 den=10.5 1 900 92234 0.02543 

 92235 3.00 

 92236 0.01380 

 92238 96.9608 end 

'clad 

zirc2 2 1 573 end 

'water coolant 

h2o 3 den=0.5 1 552 end 

'gap 

he-4 4 den=0.82e-3 1 573 end 

'fuel 2 cell type (with gadolinia 3.5%) 

uo2 5 den=10.5 0.965 900 92234 0.02543 

 92235 3.00 

 92236 0.01380 

 92238 96.9608 end 

arbmgd 10.5 2 0 1 0 64000 2 8016 3 5 0.035 900 end 

'clad 

zirc2 6 1 573 end 

'water coolant 

h2o 7 den=0.5 1 552 end 

'gap 

he-4 8 den=0.82e-3 1 573 end 

'channel box 

zirc4 9 1 552 end 

'water moderator 

h2o 10 den=0.75 1 552 end 

end comp 

'======== 

read celldata 

 latticecell squarepitch pitch=1.24 3 

 fueld=0.819 1 

 cladd=0.962 2 

 gapd=0.836 4 end 

 latticecell squarepitch pitch=1.24 7 

 fueld=0.819 5 

 cladd=0.962 6 

 gapd=0.836 8 end 

end celldata 

'======== 

read depletion 

 1 5 

end depletion 

'======== 

read burndata 

 power=40.0 burn=1e-15 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

 power=40.0 burn=75 down=0 end 

end burndata 

'======== 
B-27 

read model 

BWR SVEA-100 

'-------- 

read parm 

 fillmix=10 epseigen=1e-4 epsinner=1e-4 

 epsouter=1e-4 cmfd=yes xycmfd=4 echo=yes 

end parm 

'-------- 

read materials 

 1 1 ! fuel ! end 

 2 1 ! clad ! end 

 3 2 ! coolant ! end 

 4 0 ! gap ! end 

 5 1 ! fuel-Gd (3.5%) ! end 

 9 1 ! channel box ! end 

10 2 ! moderator ! end 

end materials 

'-------- 

read geom 

unit 1 

com='fuel pin' 

 cylinder 10 .4095 

 cylinder 20 .418 

 cylinder 30 .481 

 cuboid 40 4p0.62 

 media 1 1 10 

 media 4 1 20 -10 

 media 2 1 30 -20 

 media 3 1 40 -30 

 boundary 40 4 4 

unit 2 

com='fuel wih Gd 3.5%' 

 cylinder 10 .4095 

 cylinder 20 .418 

 cylinder 30 .481 

 cuboid 40 4p0.62 

 media 5 1 10 

 media 4 1 20 -10 

 media 2 1 30 -20 

 media 3 1 40 -30 

 boundary 40 4 4 

unit 3 

com='sub-assembly NE' 

 cuboid 31 4p3.28 

 cuboid 32 3.28 -3.39 3.28 -3.39 

 array 1 31 place 3 3 0. 0. 

 media 3 1 31 

 media 9 1 32 -31 

 boundary 32 20 20 

unit 4 

com='sub-assembly NW' 

 cuboid 41 4p3.28 

 cuboid 42 3.39 -3.28 3.28 -3.39 

 array 2 41 place 3 3 0. 0. 

 media 3 1 41 

 media 9 1 42 -41 

 boundary 42 20 20 

unit 5 

com='sub-assembly SW' 

 cuboid 51 4p3.28 

 cuboid 52 3.39 -3.28 3.39 -3.28 

 array 3 51 place 3 3 0. 0. 

 media 3 1 51 

 media 9 1 52 -51 

 boundary 52 20 20 

unit 6 

com='sub-assembly SE' 

 cuboid 61 4p3.28 

 cuboid 62 3.28 -3.39 3.39 -3.28 

 array 4 61 place 3 3 0. 0. 

 media 3 1 61 

 media 9 1 62 -61 

 boundary 62 20 20 

global unit 7 

com='assembly' 

 cuboid 71 4p6.87 

 cuboid 72 4p6.98 

 cuboid 73 4p7.725 

 hole 3 origin x=3.59 y=3.59 

 hole 4 origin x=-3.59 y=3.59 

 hole 5 origin x=-3.59 y=-3.59 
B-28 

 hole 6 origin x=3.59 y=-3.59 

 media 10 1 71 

 media 9 1 72 -71 

 media 10 1 73 -72 

 boundary 73 20 20 

end geom 

'-------- 

read array 

 ara=1 nux=5 nuy=5 typ=cuboidal 

 fill 

 1 1 1 1 1 

 1 1 1 2 1 

 1 1 1 1 1 

 1 2 1 1 1 

 1 1 1 1 1 

 end fill 

 ara=2 nux=5 nuy=5 typ=cuboidal 

 fill 

 1 1 1 1 1 

 1 1 1 1 1 

 1 1 2 1 1 

 1 1 1 1 1 

 1 1 1 1 1 

 end fill 

 ara=3 nux=5 nuy=5 typ=cuboidal 

 fill 

 1 1 1 1 1 

 1 1 1 2 1 

 1 1 1 1 1 

 1 2 1 1 1 

 1 1 1 1 1 

 end fill 

 ara=4 nux=5 nuy=5 typ=cuboidal 

 fill 

 1 1 1 1 1 

 1 1 1 1 1 

 1 1 2 1 1 

 1 1 1 1 1 

 1 1 1 1 1 

 end fill 

end array 

'-------- 

read bounds 

 all=refl 

end bounds 

'-------- 

end data 

'======== 

end 

*

 Trace nuclides need to be included in fuel composition, as for the 7 × 7 assembly; not shown here for brevity. 
C-1 

APPENDIX C

SPENT FUEL ASSEMBLY BENCHMARK DATA 

COMPARISON OF TRITON AND SAS2H RESULTS 
C-2 
C-3 

Table C.1 Comparison of TRITON and SAS2H decay heat results for selected Swedish PWR fuel ass

Reactor Assembly ID Assembly design Enrichment (wt % 235U) Final burnup (MWd/MTU) Cooling time (days) Measured (W) TRITON Calculated (W) SAS2H Calculated (W) 

Ringhals 2 C01 15×15 3.1 36,688 8468 415.8 439.3 430.4 Ringhals 2 C12 15×15 3.1 36,385 8403 410.3 434.4 425.4 Ringhals 2 C20 15×15 3.1 35,720 6950 415.8 451.9 440.1 Ringhals 2 C20 15×15 3.1 35,720 6951 426 451.9 440.1 Ringhals 2 C20 15×15 3.1 35,720 6952 428.9 451.8 440.0 Ringhals 2 C20 15×15 3.1 35,720 6959 435.7 451.7 439.9 Ringhals 2 D27 15×15 3.3 39,676 7669 456.1 473.6 464.0 Ringhals 2 D38 15×15 3.3 39,403 8005 442.3 461.7 453.5 Ringhals 2 E38 15×15 3.2 33,973 7999 376.3 390 383.1 Ringhals 2 E38 15×15 3.2 33,973 8000 374.3 390 383.1 Ringhals 2 E40 15×15 3.2 34,339 8075 381.3 394.2 387.2 Ringhals 2 F14 15×15 3.2 34,009 7722 381.8 398.9 393.2 Ringhals 2 F21 15×15 3.2 36,273 7376 420.9 433.7 428.0 Ringhals 2 F25 15×15 3.2 35,352 7725 396.7 418 411.6 Ringhals 2 F32 15×15 3.2 50,962 5860 692 732.7 713.1 Ringhals 2 G11 15×15 3.2 35,463 6990 416.4 428.5 423.5 Ringhals 2 G23 15×15 3.2 35,633 6984 420.6 436.6 430.8 Ringhals 2 I09 15×15 3.2 40,188 5849 507.9 535.4 530.5 Ringhals 2 I24 15×15 3.2 34,294 6601 410.1 419 414.7 Ringhals 2 I25 15×15 3.2 36,859 6198 445.8 467.3 463.3 Ringhals 3 0E2 17×17 3.1 41,628 5823 587.9 609 598.0 Ringhals 3 0E6 17×17 3.1 35,993 5829 487.8 505.6 507.6 Ringhals 3 1C2 17×17 3.1 33,318 6559 417.7 438.3 436.3 Ringhals 3 1E5 17×17 3.1 34,638 5818 468.8 485.2 485.5 Ringhals 3 2A5 17×17 2.1 20,107 7297 233.8 246.3 245.3 Ringhals 3 2C2 17×17 3.1 36,577 6550 466.5 491 487.2 
C-4 

Table C.1 Comparison of TRITON and SAS2H decay heat results for selected Swedish PWR fuel ass

(continued) 

Reactor Assembly ID Assembly design Enrichment (wt % 235U) Final burnup (MWd/MTU) Cooling time (days) Measured (W) TRITON Calculated (W) SAS2H Calculated (W) 

Ringhals 3 3C1 17×17 3.1 36,572 6545 470.2 489.9 483.0 Ringhals 3 4C4 17×17 3.1 33,333 6572 422 438.9 436.4 Ringhals 3 5A3 17×17 2.1 19,699 6972 237.7 244.2 243.8 Ringhals 3 5A3 17×17 2.1 19,699 6975 236.7 244.2 243.8 Ringhals 3 5A3 17×17 2.1 19,699 6977 243.4 244.2 239.8 Ringhals 3 5A3 17×17 2.1 19,699 7291 230.9 240.7 239.6 

a (SAS2H / TRITON) – 1 × 100%. 
C-5 

Table C.2 Comparison of TRITON and SAS2H decay heat results for selected Swedish BWR fuel ass

Reactor Assembly ID Assembly design Enrichment (wt % 235U) Final burnup (MWd/MTU) Cooling time (days) Measured (W) TRITON Calculated (W) SAS2H Calculate

(W) 

Ringhals 1 1177 8×8 2.64 36,242 6690 177.9 174 177.29 Ringhals 1 1186 8×8 2.64 30,498 6674 140.8 142.5 144.12 Ringhals 1 6423 8×8 2.9 35,109 5669 174.2 175.7 174.74 Ringhals 1 6432 8×8 2.89 36,861 5422 185.5 187 185.87 Ringhals 1 6432 8×8 2.89 36,861 5424 189.5 186.9 185.87 Ringhals 1 6432 8×8 2.89 36,861 5670 184.4 184.1 183.15 Ringhals 1 6432 8×8 2.89 36,861 5680 182.8 184 183.05 Ringhals 1 6432 8×8 2.89 36,861 5681 185.2 184 182.95 Ringhals 1 6432 8×8 2.89 36,861 5687 181.6 183.9 182.95 Ringhals 1 6432 8×8 2.89 36,861 5688 182 183.9 182.85 Ringhals 1 6454 8×8 2.9 37,236 6395 186.3 181.1 179.69 Ringhals 1 8327 8×8 2.9 37,851 4600 196.9 201.2 199.61 Ringhals 1 8331 8×8 2.91 35,903 5291 187 183.7 182.39 Ringhals 1 8332 8×8 2.89 34,977 5690 168.1 172.5 171.33 Ringhals 1 8338 8×8 2.91 34,830 5695 169.5 172.5 170.46 Forsmark 1 3838 8×8 2.09 25,669 4170 126.8 131.7 132.72 Forsmark 1 3838 8×8 2.09 25,669 4171 125.9 131.7 132.71 Forsmark 1 KU0100 8×8−2 2.98 34,193 4893 185.3 179.3 178.61 Forsmark 2 5535 8×8 2.1 19,944 5634 84.6 92.5 92.61 Oskarshamn 2 1377 8×8 2.2 14,546 9750 56.2 59.4 59.26 Oskarshamn 2 1389 8×8 2.2 19,481 8171 83.9 83.4 84.03 Oskarshamn 2 1546 8×8 2.2 24,470 7455 108.1 110.9 111.00 Oskarshamn 2 1696 8×8 2.2 20,870 7411 92.4 93.8 92.56 Oskarshamn 2 1704 8×8 2.2 19,437 7808 84 84.1 84.43 Oskarshamn 2 2995 8×8 2.7 29,978 8211 130.5 131.6 131.87 Oskarshamn 2 6350 8×8 2.88 27,675 6754 126.9 129.1 129.22 
C-6 

Table C.2 Comparison of TRITON and SAS2H decay heat results for selected Swedish BWR fuel ass

(continued) 

Reactor Assembly ID Assembly design Enrichment (wt % 235U) Final burnup (MWd/MTU) Cooling time (days) Measured (W) TRITON Calculated (W) SAS2H Calculate

(W) 

Oskarshamn 2 6350 8×8 2.88 27,675 6755 129.4 129.2 129.22 Oskarshamn 3 12078 8×8 2.58 25,160 5611 120.2 125.6 126.81 Barsebäck 1 9329 8×8 2.92 41,127 5371 222.8 221.7 222.08 Barsebäck 1 9329 8×8 2.92 41,127 5373 224.4 221.7 221.98 Barsebäck 1 9329 8×8 2.92 41,127 5542 218.7 219.2 219.65 Barsebäck 1 10288 8×8 2.95 35,218 5534 185.8 183.8 183.15 Barsebäck 2 14076 8×8 3.15 40,010 4177 240.3 235.7 232.83 Forsmark 1 KU0269 9×9−5 2.94 35,113 4903 192.7 191.9 191.26 Forsmark 1 KU0278 9×9−5 2.94 35,323 4595 195.4 194.1 193.81 Forsmark 1 KU0282 9×9−5 2.94 37,896 4574 218.5 211.9 211.93 Forsmark 2 11494 SVEA-64 2.92 32,431 5618 166 166 164.92 Forsmark 2 11495 SVEA-64 2.91 32,431 5593 167.6 166.9 164.89 Forsmark 2 13775 SVEA-64 2.85 32,837 4543 178.4 179 177.87 Oskarshamn 2 12684 SVEA-64 2.9 46,648 4519 282.7 280.1 276.82 Forsmark 3 13847 SVEA-100 2.77 31,275 4871 170.3 168.5 168.60 Forsmark 3 13847 SVEA-100 2.77 31,275 4871 169.6 168.5 168.60 Forsmark 3 13848 SVEA-100 2.77 31,275 4882 170.7 168.5 168.40 Oskarshamn 3 13628 SVEA-100 2.71 35,619 4581 194 196.5 196.02 Oskarshamn 3 13630 SVEA-100 2.71 40,363 4554 235.7 232.9 229.83 

a (SAS2H / TRITON) – 1 × 100%. 

NUREG/CR-6972 

ORNL/TM-2008/015 

December 2009