NUREG/CR-6971 

ORNL/TM-2008/016 

Spent Fuel Decay Heat 

Measurements Performed at 

the Swedish Central Interim 

Storage Facility 

Office of Nuclear Regulatory Research 
AVAILABILITY OF REFERENCE MATERIALS 

IN NRC PUBLICATIONS

NRC Reference Material

As of November 1999, you may electronically access 

NUREG-series publications and other NRC records at 

NRC’s Public Electronic Reading Room at 

http://www.nrc.gov/reading-rm.html. Publicly released 

records include, to name a few, NUREG-series 

publications; Federal Register notices; applicant, 

licensee, and vendor documents and correspondence; 

NRC correspondence and internal memoranda; bulletins 

and information notices; inspection and investigative 

reports; licensee event reports; and Commission papers 

and their attachments. 

NRC publications in the NUREG series, NRC 

regulations, and Title 10, Energy, in the Code of 

Federal Regulations may also be purchased from one 

of these two sources: 

1. The Superintendent of Documents 

 U.S. Government Printing Office 

 P.O. Box SSOP 

 Washington, DC 20402–0001 

 Internet: bookstore.gpo.gov 

 Telephone: 202–512–1800 

 Fax: 202-512-2250 

2. The National Technical Information Service 

Springfield, VA 22161–0002 

 www.ntis.gov 

 1–800–553–6847 or, locally, 703–605–6000 

A single copy of each NRC draft report for comment is 

available free, to the extent of supply, upon written 

request as follows: 

Address: Office of the Chief Information Officer, 

 Reproduction and Distribution 

 Services Section 

 U.S. Nuclear Regulatory Commission 

 Washington, DC 20555–0001 

E-mail: DISTRIBUTION@nrc.gov 

Facsimile: 301–415–2289 

Some publications in the NUREG series that are 

posted at NRC’s Web site address 

http://www.nrc.gov/reading-rm/doc-collections/nuregs

are updated periodically and may differ from the last 

printed version. Although references to material found 

on a Web site bear the date the material was accessed, 

the material available on the date cited may 

subsequently be removed from the site. 

Non-NRC Reference Material

Documents available from public and special technical 

libraries include all open literature items, such as 

books, journal articles, and transactions, Federal 

Register notices, Federal and State legislation, and 

congressional reports. Such documents as theses, 

dissertations, foreign reports and translations, and 

non-NRC conference proceedings may be purchased 

from their sponsoring organization. 

Copies of industry codes and standards used in a 

substantive manner in the NRC regulatory process are 

maintained at— 

 The NRC Technical Library 

 Two White Flint North 

 11545 Rockville Pike 

 Rockville, MD 20852–2738 

These standards are available in the library for 

reference use by the public. Codes and standards are 

usually copyrighted and may be purchased from the 

originating organization or, if they are American 

National Standards, from— 

 American National Standards Institute 

 11 West 42nd Street 

 New York, NY 10036–8002 

 www.ansi.org 

 212–642–4900 

DISCLAIMER: This report was prepared as an account of work sponsored by an agency of the U.S. Government. 

Neither the U.S. Government nor any agency thereof, nor any employee, makes any warranty, expressed or 

implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any 

information, apparatus, product, or process disclosed in this publication, or represents that its use by such third 

party would not infringe privately owned rights.

Legally binding regulatory requirements are stated 

only in laws; NRC regulations; licenses, including 

technical specifications; or orders, not in NUREGseries publications. The views expressed in 

contractor-prepared publications in this series are 

not necessarily those of the NRC. 

The NUREG series comprises (1) technical and 

administrative reports and books prepared by the 

staff (NUREG/XXXX) or agency contractors 

(NUREG/CR-XXXX), (2) proceedings of 

conferences (NUREG/CP-XXXX), (3) reports 

resulting from international agreements 

(NUREG/IA-XXXX), (4) brochures (NUREG/BRXXXX), and (5) compilations of legal decisions and 

orders of the Commission and Atomic and Safety 

Licensing Boards and of Directors' decisions under 

Section 2.206 of NRC's regulations (NUREG0750). 
NUREG/CR-6971 

ORNL/TM-2008/016 

Spent Fuel Decay Heat 

Measurements Performed at 

the Swedish Central Interim 

Storage Facility 

Manuscript Completed: June 2009 

Date Published: February 2010 

Prepared by 

B.D. Murphy and I.C. Gauld 

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 

Calorimeter measurements of full-length spent fuel assemblies have been performed at the Swedish 

Central Interim Storage Facility for Spent Fuel (CLAB) located in Oskarshamn, Sweden. These 

measurements provide an important and unique database for the validation of computer code predictions 

of decay heat for modern fuel types and assembly designs. The measured assemblies include boilingwater reactor 8 × 8, 9 × 9, SVEA-64, and SVEA-100 assemblies, and pressurized-water reactor 15 × 15 

and 17 × 17 assemblies. This report compiles the assembly design and fuel data, reactor operating history 

data, reactor conditions, and the results of the decay heat measurements that have been performed for 

intact spent fuel assemblies. The data are presented in sufficient detail to allow independent code 

validation studies to be performed.
iv 
v 

TABLE OF CONTENTS 

Page

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

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

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

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

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

2 FACILITY DESCRIPTION................................................................................................................ 3

3 REACTOR DESCRIPTIONS............................................................................................................. 7

4 ASSEMBLY DESIGN DATA.......................................................................................................... 15

4.1 Overview of Assembly Designs .............................................................................................. 15

4.2 17 × 17 PWR Assemblies........................................................................................................ 16

4.3 15 × 15 PWR Assemblies........................................................................................................ 16

4.4 8 × 8 BWR Assemblies ........................................................................................................... 19

4.5 9 × 9 BWR Assemblies ........................................................................................................... 19

4.6 SVEA Assemblies ................................................................................................................... 19

5 FUEL DATA..................................................................................................................................... 25

5.1 PWR 15 × 15 Assemblies........................................................................................................ 25

5.2 PWR 17 × 17 Assemblies........................................................................................................ 25

5.3 BWR 8 × 8 Assemblies ........................................................................................................... 25

5.4 BWR 9 × 9 Assemblies ........................................................................................................... 26

5.5 SVEA Assemblies ................................................................................................................... 26

6 POWER HISTORY DATA .............................................................................................................. 33

6.1 Assembly Burnup Data ........................................................................................................... 33

6.2 Assembly Burnup Profile Data................................................................................................ 40

6.3 Burnup-Profile Trends............................................................................................................. 50

7 MODERATOR DATA ..................................................................................................................... 53

7.1 Boron Levels ........................................................................................................................... 53

7.2 Void Profiles ........................................................................................................................... 54

7.3 Pressure-Density-Temperature................................................................................................ 54

7.4 Trends in Void-Profile Data .................................................................................................... 61

8 SUMMARY ...................................................................................................................................... 65

9 REFERENCES.................................................................................................................................. 67
vi 

TABLE OF CONTENTS (continued) 

Page

APPENDIX A: ASSEMBLY DIAGRAMS ............................................................................................A-1

A.1 PWR Assembly Configurations ............................................................................A-2

A.2 BWR Assembly Configurations...........................................................................A-4

APPENDIX B: MEASURED DECAY HEAT VALUES....................................................................... B-1
vii 

LIST OF FIGURES 

 Page

Figure 1 Spent fuel assembly calorimeter at the Swedish CLAB facility.............................................. 4 

Figure 2 Spent fuel assembly calorimeter .............................................................................................. 5 

Figure 3 View of calorimeter controller and digital readout monitor .................................................... 6 

Figure 4 PWR burnup profiles for Ringhals 3 reactor assemblies ....................................................... 50 

Figure 5 BWR burnup profile for Barsebäck 2 assembly 14076 ......................................................... 51 

Figure 6 BWR burnup profiles for selected Oskarshamn 2 assemblies ............................................... 51 

Figure 7 Void fraction profile for Barsebäck 1 assemblies.................................................................. 61

Figure 8 Void fraction profile for Oskarshamn 2 assemblies .............................................................. 62 

Figure 9 Maximum void fraction vs. assembly average power ........................................................... 63 

Figure A.1 17 × 17 PWR assembly layout.............................................................................................A-2

Figure A.2 15 × 15 PWR assembly layout.............................................................................................A-3

Figure A.3 8 × 8 BWR assembly layout ................................................................................................A-4

Figure A.4 9 × 9 BWR assembly layout ................................................................................................A-5

Figure A.5 SVEA-64 assembly layout...................................................................................................A-6 

Figure A.6 SVEA-100 assembly layout.................................................................................................A-7 
viii 
ix 

LIST OF TABLES 

 Page

Table 1 Reactor operating and design data............................................................................................. 8

Table 2 Barsebäck 1 reactor operating history data ............................................................................... 9

Table 3 Barsebäck 2 reactor operating history data ............................................................................... 9

Table 4 Forsmark 1 reactor operating history data............................................................................... 10

Table 5 Forsmark 2 reactor operating history data............................................................................... 10

Table 6 Forsmark 3 reactor operating history data............................................................................... 10

Table 7 Oskarshamn 2 reactor operating history data .......................................................................... 11

Table 8 Oskarshamn 3 reactor operating history data .......................................................................... 11

Table 9 Ringhals 1 reactor operating history data................................................................................ 12

Table 10 Ringhals 2 reactor operating history data................................................................................ 12

Table 11 Ringhals 3 reactor operating history data................................................................................ 13

Table 12 Summary of measured Swedish fuel assemblies..................................................................... 15

Table 13 17 × 17 PWR assembly design data ........................................................................................ 17

Table 14 15 × 15 PWR assembly design data ........................................................................................ 18

Table 15 8 × 8 BWR assembly design data............................................................................................ 21

Table 16 9 × 9 BWR assembly design data............................................................................................ 22

Table 17 SVEA-64 assembly design data .............................................................................................. 23

Table 18 SVEA-100 assembly design data ............................................................................................ 24

Table 19 Fuel data for PWR 15 × 15 assemblies ................................................................................... 27

Table 20 Fuel data for PWR 17 × 17 assemblies ................................................................................... 28

Table 21 Fuel data for BWR 8 × 8 assemblies....................................................................................... 29

Table 22 Fuel data for BWR 9 × 9 assemblies....................................................................................... 30

Table 23 Fuel data for SVEA-64 assemblies.......................................................................................... 31

Table 24 Fuel data for SVEA-100 assemblies........................................................................................ 31

Table 25 Barsebäck 1 assembly irradiation history................................................................................ 33

Table 26 Barsebäck 2 assembly irradiation history................................................................................ 33

Table 27 Forsmark 1 assembly irradiation history ................................................................................. 34

Table 28 Forsmark 2 assembly irradiation history ................................................................................. 34

Table 29 Forsmark 3 assembly irradiation history ................................................................................. 35

Table 30 Oskarshamn 2 assembly irradiation history ............................................................................ 35
x 

LIST OF TABLES (continued) 

Page

Table 31 Oskarshamn 3 assembly irradiation history ............................................................................ 36

Table 32 Ringhals 1 assembly irradiation history .................................................................................. 37

Table 33 Ringhals 2 assembly irradiation history .................................................................................. 38

Table 34 Ringhals 3 assembly irradiation history .................................................................................. 39

Table 35 PWR 15 × 15 assembly burnup profile data............................................................................ 41

Table 36 PWR 17 × 17 assembly burnup profile data............................................................................ 43

Table 37 BWR 8 × 8 assembly burnup profile data ............................................................................... 45

Table 38 BWR 9 × 9 assembly burnup profile data ............................................................................... 47

Table 39 SVEA-64 assembly burnup profile data.................................................................................. 48

Table 40 SVEA-100 assembly burnup profile data................................................................................ 49

Table 41 Average soluble boron concentrations for Ringhals 2 and 3................................................... 53

Table 42 Void fractions for 8 × 8 assemblies......................................................................................... 55

Table 43 Void fractions for 9 × 9 assemblies......................................................................................... 57

Table 44 Void fractions for SVEA-64 assemblies ................................................................................. 58

Table 45 Void fractions for SVEA-100 assemblies ............................................................................... 59

Table 46 Water moderator densities for Forsmark assemblies............................................................... 60

Table B.1 Decay heat measurement results for PWR assemblies.......................................................... B-2

Table B.2 Decay heat measurement results for BWR assemblies ......................................................... B-4
xi 

ACKNOWLEDGMENTS 

This work was supported by the Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory 

Commission (NRC), under Project Y6517, High Burnup Source Terms for Storage. The collaboration 

was made possible under an agreement between Oak Ridge National Laboratory and the Swedish Nuclear 

Fuel and Waste Management Co., Svensk Kärnbränslehuntering AB. The agreement and technical 

cooperation were facilitated by Lennart Agrenius, of Agrenius Engineering Co. (Sweden). The authors 

acknowledge the review of the manuscript by G. Ilas and G. Radulescu at Oak Ridge National 

Laboratory, the formatting by W. C. Carter and D. Weaver, editing by C. C. Southmayd, and 

administrative support by A. C. Alford. 
xii 
1 

1 INTRODUCTION 

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

has initiated a large experimental program to measure decay heat from boiling-water reactor (BWR) and 

pressurized-water reactor (PWR) assemblies residing at the Swedish Central Interim Storage Facility for 

Spent Nuclear Fuel, CLAB, located outside Oskarshamn, Sweden.1

 Oak Ridge National Laboratory 

(ORNL), under a collaboration agreement with SKB, has provided technical support to the program in the 

areas of calorimeter design, uncertainty assessment, evaluation of the fuel design data and experimental 

data, and validation of computational methods using the ORNL SCALE code system2

 to predict assembly 

decay heat. The decay heat measurements performed at CLAB are being used to extend the validation 

database, beyond that currently available, to include modern-design assemblies and high burnup fuel. 

The measurements are being used by ORNL to validate computer code predictions of decay heat for a 

wide range of fuel designs and support the technical basis for expanding the U.S. Nuclear Regulatory 

Commission (NRC) Regulatory Guide RG 3.54, Spent Fuel Heat Generation in an Independent Spent 

Fuel Storage Installation.3

 

This report presents a compilation of the experimental measurements, fuel and assembly design data, fuel 

irradiation data, and reactor operating information necessary to perform benchmark calculations. A more 

extensive report on the measurements and assembly data has been published separately by SKB.4

 

The assemblies described in this report include 8 × 8, 9 × 9, SVEA-64, SVEA-100, 15 × 15, and 17 × 17 

designs. This report is restricted to include only the measurements of intact fuel assemblies. Those 

assemblies that were rebuilt (fuel rods either removed or reconfigured) are not included at this time 

because insufficient information was available on the details of the reconstitution to allow the assemblies 

to be considered for use in code validation. A total of 34 BWR assemblies and 30 PWR assemblies are 

described in this report. Multiple measurements were performed on some assemblies. All calorimeter 

measurements documented in this report were performed at the CLAB facility between June 2003 and 

June 2004. 

The fuel data and measurement information documented in this report were supplied by SKB. Reactor 

operating history data, assembly burnup data by reactor cycle, axial void profiles, and other reactor data is 

based on information provided by Swedish nuclear power plants. Detailed information was not available 

for all assemblies at the time this study was performed. For these assemblies, appropriate values were 

estimated based on the data for similar assemblies with more detailed descriptions. Any information in 

this report that has been estimated or inferred from other assemblies is indicated. 
2 
3 

2 FACILITY DESCRIPTION 

All fuel assembly calorimeter measurements described in this report were performed at the Swedish 

CLAB spent fuel storage facility located near the site of the Oskarshamn Nuclear Power Station owned by 

OKG, in Sweden. The CLAB facility is operated by OKG for SKB. Spent nuclear fuel from all Swedish 

reactors is stored at CLAB in pools located 30 m below ground, with a capacity of 5,000 tonnes of 

uranium (about 20,000 BWR assemblies and 2,500 PWR assemblies). 

To support the characterization of the spent fuel inventory at CLAB, SKB has undertaken a project to 

construct a spent fuel assembly calorimeter and perform measurements for use in validating computer 

code predictions of decay heat. Measurements will be repeated on selected fuel assemblies 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. The experimental program being conducted by SKB is being used 

to validate the computer code and data used to support the safety analysis of the final repository for 

radioactive waste to be constructed in Sweden. 

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

Operations.5,6 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 centered and vertical in the calorimeter. Temperatures within the calorimeter are measured by 

16 PT-100 type sensors: 8 sensors in the water inside the vessel; 2 sensors on the inside and outside 

surfaces of the calorimeter; and 2 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, which is used to correct the calorimeter data for gamma energy loss. 

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

establish the thermal output of the assemblies. 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 the measured energy loss from gamma rays escaping from the 

calorimeter. The calorimeter is shown at poolside and raised to show the assembly loading hatch of the 

calorimeter above the water level in Figures 1 and 2. The digital controller and readout monitor are 

shown in Figure 3. 

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

95 % 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 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 estimated7

 using the results of repeat measurements for six of the BWR 

assemblies and four of the PWR assemblies. The standard deviation for this group of BWR and PWR 

assemblies was found 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; 

however, the population of assemblies is small in both cases. The uncertainty analysis described includes 

only random sources of uncertainty and does not address potential sources of systematic bias. A more 
4 

detailed description of the analysis of measurements and evaluation of measurement uncertainty is 

documented in Ref. 7. 

Figure 1 Spent fuel assembly calorimeter at the Swedish CLAB facility 
5 

Figure 2 Spent fuel assembly calorimeter (side view) 
6 

Figure 3 View of calorimeter controller (top) and digital readout monitor (bottom) 
7 

3 REACTOR DESCRIPTIONS 

The BWR and PWR assemblies that were measured were irradiated in ten different Swedish reactors. 

Some general reactor information and other data relevant to the decay heat simulations are listed in 

Table 1. Also listed in the table are the number of assemblies in the core and (where available) the total 

mass of uranium in the core. 

The BWR assemblies described in this report are from the following nuclear power stations: 

Barsebäck 1: Two 8 × 8 assemblies. 

Barsebäck 2: One 8 × 8 assembly. 

Forsmark 1: Two 8 × 8 assemblies and three 9 × 9 assemblies. 

Forsmark 2: One 8 × 8 assembly and three SVEA-64 assemblies. 

Forsmark 3: Two SVEA-100 assemblies. 

Oskarshamn 2: Seven 8 × 8 assemblies and one SVEA-64 assembly. 

Oskarshamn 3: One 8 × 8 assembly and two SVEA-100 assemblies. 

Ringhals 1: Nine 8 × 8 assemblies. 

The PWR assemblies are from the following nuclear power stations: 

Ringhals 2: Sixteen 15 × 15 assemblies. 

Ringhals 3: Fourteen 17 × 17 assemblies. 

Tables 2–11 list the operating data for each reactor. Most important among these are the beginning and 

ending dates of all cycles of the assemblies in the study. These dates were used to determine the cycle 

lengths and downtime between cycles. 

An important operating parameter for the PWR plants is the soluble boron concentration in the moderator. 

This information was not provided for either the Ringhals 2 or 3 plants. Based on PWR reactor data 

compiled in previous validation studies,7–11 typical cycle-averaged boron levels may be assumed. A 

parametric study found that the predicted decay heat was not very sensitive to the boron level for the 

cooling times of interest, indicating that average boron levels are adequate. 
8

Table 1 Reactor operating and design dataa

Reactor Type Pressure (MPa) Inlet temperature (K) Outlet temperature (K) Assembly pitch (cm) Assemblies in core U 

in 

(

Barsebäck 1 BWR 6.86 545 559 15.30 444 

Barsebäck 2 BWR 6.86 545 559 15.30 444 Forsmark 1 BWR 6.86 547 559 15.38 676 1Forsmark 2 BWR 6.86 547 559 15.40 676 1Forsmark 3 BWR 6.86 551 559 15.45 700 1Oskarshamn 2 BWR 6.86 547 559 15.30 444 Oskarshamn 3 BWR 6.86 550 561 15.50 700 1Ringhals 1 BWR 6.86 545 559 15.33 648 1

Ringhals 2 PWR 15.1 559 595 21.50 157 Ringhals 3 PWR 15.3 557 595 21.50 157 

a Data for the reactor operating pressures, inlet and outlet temperatures, and assembly pitch, were obtained from the 

Reactor Handbook, 2002 (Ref. 12). 
9 

Table 2 Barsebäck 1 reactor operating history data 

Cycle Start 

date 

Stop 

date 

Cycle length 

(days) 

Downtime 

(days) 

8 8/6/1983a 6/28/1984 327 28 

9 7/26/1984 8/7/1985 377 26 

10 9/2/1985 7/17/1986 318 32 

11 8/18/1986 7/1/1987 317 28 

12 7/29/1987 9/17/1988 416 

a

 Dates given as month/day/year. 

Table 3 Barsebäck 2 reactor operating history data 

Cycle Start 

date 

Stop 

date 

Cycle length 

(days) 

Downtime 

(days) 

9 9/19/1987 7/6/1988 291 32 

10 8/7/1988 9/8/1989 397 20 

11 9/28/1989 7/11/1990 286 38 

12 8/18/1990 9/6/1991 384 16 

13 9/22/1991 7/2/1992 284 37 
10 

Table 4 Forsmark 1 reactor operating history data 

Cycle Start 

date 

Stop 

date 

Cycle 

length 

(days) 

Downtime 

(days) 

1A 4/23/1980 6/25/1981 428 52 

1Ba 8/16/1981 6/19/1982 307 67 

2 8/25/1982 7/15/1983 324 21 

3 8/5/1983 7/13/1984 343 24 

4 8/6/1984 5/31/1985 298 38 

5 7/8/1985 7/4/1986 361 18 

6 7/22/1986 7/31/1987 374 19 

7 8/19/1987 6/10/1988 296 30 

8 7/10/1988 7/14/1989 369 35 

9 8/18/1989 8/17/1990 364 30 

10 9/16/1990 5/24/1991 250 27 

a

 Reactor cycle simulated in multiple phases. 

Table 5 Forsmark 2 reactor operating history data 

Cycle Start 

date 

Stop 

date 

Cycle 

length 

(days) 

Downtime 

(days) 

1A 12/1/1980 5/1/1982 516 109 

1B 8/18/1982 5/20/1983 275 22 

2 6/11/1983 5/25/1984 349 49 

3 7/13/1984 7/5/1985 357 21 

4 7/26/1985 7/25/1986 364 15 

5 8/9/1986 5/31/1987 295 20 

6 6/20/1987 7/15/1988 391 20 

7 8/4/1988 8/18/1989 379 17 

8 9/4/1989 5/18/1990 256 25 

9 6/12/1990 7/12/1991 395 19 

Table 6 Forsmark 3 reactor operating history data 

Cycle Start 

date 

Stop 

date 

Cycle 

length 

(days) 

Downtime 

(days) 

2 9/2/1986 7/10/1987 311 30 

3 8/9/1987 8/12/1988 369 24 

4 9/5/1988 6/9/1989 277 34 

5 7/13/1989 7/13/1990 365 16 
11 

Table 7 Oskarshamn 2 reactor operating history data 

Cycle Start 

date 

Stop 

date 

Cycle 

length 

(days) 

Downtime 

(days) 

1 10/2/1974 4/16/1976 562 77 

2 7/2/1976 5/13/1977 315 86 

3 8/7/1977 6/24/1978 321 71 

4 9/3/1978 7/20/1979 320 41 

5 8/30/1979 7/12/1980 317 39 

6 8/20/1980 7/15/1981 329 46 

7 8/30/1981 7/23/1982 327 31 

8 8/23/1982 8/19/1983 361 31 

9 9/19/1983 7/1/1984 286 54 

10 8/24/1984 6/7/1985 287 36 

11 7/13/1985 8/16/1986 399 42 

12 9/27/1986 7/31/1987 307 35 

13 9/4/1987 8/20/1988 351 21 

14 9/10/1988 8/5/1989 329 27 

15 9/1/1989 8/10/1990 343 62 

16 10/11/1990 8/2/1991 295 

Table 8 Oskarshamn 3 reactor operating history data 

Cycle Start 

date 

Stop 

date 

Cycle 

length 

(days) 

Downtime 

(days) 

1 3/18/1985 7/4/1986 473 22 

2 7/26/1986 7/3/1987 342 23 

3 7/26/1987 7/8/1988 348 37 

4 8/14/1988 12/10/1988 118 12 

4B 12/22/1988 6/7/1989 167 13 

5 6/20/1989 6/23/1990 368 39 

6 8/1/1990 6/24/1991 327 
12 

Table 9 Ringhals 1 reactor operating history data 

Cycle Start 

date 

Stop 

date 

Cycle 

length 

(days) 

Downtime 

(days) 

2 10/13/1978 7/12/1979 272 80 

3 9/30/1979 7/3/1980 277 97 

4 10/8/1980 7/30/1981 295 99 

5 11/6/1981 7/22/1982 258 49 

6 9/9/1982 6/18/1983 282 104 

7 9/30/1983 7/13/1984 287 46 

8 8/28/1984 8/2/1985 339 31 

9 9/2/1985 8/15/1986 347 55 

10 10/9/1986 8/21/1987 316 34 

11 9/24/1987 8/6/1988 317 36 

12 9/11/1988 9/15/1989 369 26 

13 10/11/1989 8/4/1990 297 26 

14 8/30/1990 8/9/1991 344 28 

Table 10 Ringhals 2 reactor operating history data 

Cycle Start 

date 

Stop 

date 

Cycle 

length 

(days) 

Downtime 

(days) 

1 6/19/1974 4/13/1977 1029 85 

2 7/7/1977 3/31/1978 267 56 

3 5/26/1978 4/3/1979 312 83 

4 6/25/1979 4/1/1980 281 78 

5 6/18/1980 4/4/1981 290 80 

6 6/23/1981 5/6/1982 317 84 

7 7/29/1982 4/28/1983 273 91 

8 7/28/1983 4/13/1984 260 90 

9 7/12/1984 4/4/1985 266 71 

10 6/14/1985 4/30/1986 320 63 

11 7/2/1986 4/25/1987 297 54 

12 6/18/1987 5/12/1988 329 
13 

Table 11 Ringhals 3 reactor operating history data 

Cycle Start 

date 

Stop 

date 

Cycle 

length 

(days) 

Downtime 

(days) 

1A 7/29/1980 6/2/1983 1038 104 

1B 9/14/1983 5/11/1984 240 74 

2 7/24/1984 5/25/1985 305 47 

3 7/11/1985 5/30/1986 323 49 

4 7/18/1986 6/18/1987 335 47 

5 8/4/1987 7/7/1988 338 38 
14 
15 

4 ASSEMBLY DESIGN DATA 

4.1 Overview of Assembly Designs

Swedish nuclear fuel assemblies are manufactured by Westinghouse Nuclear Fuel Operations, located in 

Västerås, Sweden. The fuel assemblies selected for the decay heat measurements at CLAB included fuels 

from the Swedish Barsebäck, Forsmark, Oskarshamn, and Ringhals Nuclear Power Stations. The 

assemblies include a broad range of designs: 8 × 8, 9 × 9, SVEA-64 (8 × 8), SVEA-100 (10 × 10), 

15 × 15, and 17 × 17. The total number of BWR and PWR assembly measurements included in this 

report was 34 and 30, respectively. Multiple decay heat measurements were made on some assemblies. 

A summary of the measured assemblies described in this report is given in Table 12; this table includes 

the enrichment range and maximum burnup of the assemblies. 

Table 12 Summary of measured Swedish fuel assemblies 

Reactor Assembly 

design 

Enrichments 

(wt % 235U) 

Number of 

assemblies 

measured 

Maximum burnup 

(MWd/MTU) 

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 

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 29,978 

Oskarshamn 2 SVEA-64 2.902 1 46,648 

Oskarshamn 3 8 × 8 2.577 1 25,160 

Oskarshamn 3 SVEA-100 2.711 2 40,363 

Ringhals 1 8 × 8 2.640, 2.911 9 37,851 

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

Ringhals 3 17 × 17 2.100 – 3.103 14 41,628 

The following sections document the assembly design details for each type of assembly measured at 

CLAB and included in this report. The designs for reconstituted assemblies are not included. In general, 

more than one set of design data exists for each assembly type. Each type of assembly is listed in a 

separate table column (identified by Type). Many of the assembly geometry design specifications are 

common to several versions. The fuel description, burnable poison rod exposure, etc. is given separately 

in Section 5. 
16 

4.2 17 × 17 PWR Assemblies

The design data for the 17 × 17 assembly configuration are listed in Table 13. The actual dimensions 

given for some assemblies were slightly different than those for other assemblies; however these 

variations were extremely small and were likely within design tolerances and of negligible neutronic 

importance. Therefore only one set of design dimensions are listed. The layout and basic design 

dimensions are illustrated in Appendix A.1. 

Data in Table 13 also include information on the spacer grid material and mass. The spacer material is 

one of the structural components of the assembly that contributes to the decay heat. Although not a major 

contributor, the structural components can be responsible for up to several percent of the decay heat. 

Essentially all of the decay heat from the structural components comes from activated cobalt in the 

Inconel spacers. 

Table 13 contains data on burnable absorber (poison) rods that are inserted in a fraction of the guide tube 

locations, typically during the first cycle of assembly irradiation. Some of the measured 17 × 17 

assemblies were exposed to burnable absorber rods. The number of guide tubes containing the absorbers 

was provided and is listed in Section 5. The burnable absorber (BA) material was assumed to be a 

borosilicate type BA (B2O3-SiO2) rod, typically used in Westinghouse 17 × 17 designs.13 The quantity of 

B2O3 in the BA rods and the design data for the rods were not provided. Based on a previous study of 

Westinghouse assembly designs,13 the BA rods are likely to be stainless steel clad absorbers that are 

annular in shape. The space internal to this annulus is a gas (not water). The design dimensions for the 

BA rods are given in Table 13. The presence (and number) of BA rods for individual assemblies is 

indicated in Section 5. 

4.3 15 × 15 PWR Assemblies

There are two different 15 × 15 assembly configurations, listed as Type 1 and 2 in Table 14. As can be 

seen from the table, the differences are related primarily to the fuel diameter and mass of spacer material. 

The layout and basic design dimensions are illustrated in Appendix A.1. 

The BA rod data are not applicable to the 15 × 15 assemblies. No BA rod exposure was reported for any 

of the 15 × 15 assemblies considered in this report. 
17 

Table 13 17 × 17 PWR assembly design data 

Design data Type 

1 

Assembly size 17 × 17 

Number of fuel rods 264 

Number of guide tubes 24 

Number of instrument tubes 1 

Fuel rod pitch (cm) 1.26 

Fuel rod diameter (cm) 0.95 

Fuel clad thickness (cm) 0.0572 

Fuel pellet diameter (cm) 0.8191 

Guide tube outer diameter (cm) 1.224 

Guide tube thickness (cm) 0.0406 

Instrument tube outer diameter (cm) 1.224 

Instrument tube thickness (cm) 0.0406 

Fuel clad material Zircaloy-4 

Guide tube material Zircaloy-4 

Instrument tube material Zircaloy-4 

Number of spacers 8 

Spacer material Inconel 718 

Spacer mass (grams per spacer) 753 

Active fuel length (cm) 365.8 

Burnable absorber rod dataa 

 Material B2O3 -SiO2

 B2O3 concentration (wt %) 12.5 

 Density (g/cm3) 2.299 

 Outer diameter (cm) 0.85344 

 Inner diameter (cm) 0.4826 

 Clad material 304 SS 

 Outer clad outer diameter (cm) 0.96774 

 Outer clad inner diameter (cm) 0.87376 

 Inner clad outer diameter (cm) 0.46101 

 Inner clad inner diameter (cm) 0.42799 

a Source: Data from Ref. 13. 
18 

Table 14 15 × 15 PWR assembly design data 

Design data Type 

1 

Type 

2 

Assembly size 15 × 15 15 × 15 

Number of fuel rods 204 204 

Number of guide tubes 20 20 

Number of instrument tubes 1 1 

Fuel rod pitch (cm) 1.43 1.43 

Fuel rod diameter (cm) 1.072 1.075 

Fuel clad thickness (cm) 0.0618 0.0725 

Fuel pellet diameter (cm) 0.929 0.911 

Guide tube outer diameter (cm) 1.387 1.389 

Guide tube thickness (cm) 0.043 0.043 

Instrument tube outer diameter (cm) 1.387 1.389 

Instrument tube thickness (cm) 0.043 0.043 

Fuel clad material Zr4 Zr4 

Guide tube material Zr4 Zr4 

Instrument tube material Zr4 Zr4 

Number of spacers 7 7 

Spacer material Inconel Inconel/Zr-4 

Spacer mass (grams per spacer) 788 720/160 

Active length (cm) 365.8 365.8 

 

 
19 

4.4 8 × 8 BWR Assemblies

Four different 8 × 8 assembly types are described in Table 15. Three of the four types use corner rods 

that have a slightly smaller diameter than those internal to the assembly. These assemblies have three 

corner rods located at each corner of the assembly. The Type 4 assemblies have fuel rods of uniform 

diameter. Minor differences in dimensions were noted for some of the assemblies. Where such 

differences were deemed to be inconsequential to the analysis of decay heat, such assemblies were 

grouped into a single assembly type to minimize the number of unique designs considered. 

The BWR assemblies specify inner and outer dimensions for a flow channel that surrounds each 

assembly, referred to here as a box. The inner diameter (flat-to-flat) of the box is slightly greater than the 

width of the fuel lattice (based on the pitch of fuel rods), resulting in a region of extra water outside of the 

fuel rods and inside of the channel box. The water within the box (with the exception of the water rods 

internal to the assembly) is considered to be in the boiling-water region, whereas the water outside the 

box is not considered as a boiling region. 

The 8 × 8 assemblies generally contain water rods. Water rods are Zircaloy tubes containing water at a 

higher density (i.e., non-boiling) than that of the boiling region to provide additional moderation in the 

fuel lattice. The fuel clad material was not provided for most assemblies, and Zircaloy-4 was generally 

assumed. From the standpoint of the decay heat analysis, it is not important whether the clad is Zircaloy2 or Zircaloy-4. Similar assumptions apply to the material in the water rods. Some assemblies also 

contained solid Zircaloy rods; referred to as spacer rods in this report. The diameter of the spacer rods is 

the same as that for the water rods. 

Many of the 8 × 8 assemblies also contained integral burnable poison rods. These poison rods contain 

gadolinium oxide (Gd2O3) mixed with the uranium oxide. These rods are not removed during the 

irradiation life of the assembly, but rather form an integral part of the assembly. The number of poison 

rods and the average poison concentration is given in Section 5. The configuration of the poison rods in 

the assembly was not provided, as this is usually considered proprietary information. A typical 

configuration for the 8 × 8 assemblies is illustrated in Appendix A.2. The arrangement of any poison rods 

in these figures is presented for illustrative purposes only, because the actual arrangement will vary for 

different assemblies. Similar considerations apply for the enrichment zoning shown in the figures. 

4.5 9 × 9 BWR Assemblies

All 9 × 9 assemblies described in this report were of the same type (i.e., same physical dimensions). The 

design data for these assemblies are listed in Table 16. The diameter of the water rods is given as being 

slightly greater than the rod pitch. The lattice configuration for a 9 × 9 assembly with five water rods is 

illustrated in Appendix A.2. The assemblies included in this report had 4 water rods and 1 Zircaloy rod. 

Fuel clad material was assumed to a Zircaloy material. 

4.6 SVEA Assemblies

Both SVEA-64 and SVEA-100 design assemblies were measured. The SVEA designs are unique from 

other 8 × 8 or 10 × 10 designs because the fuel rods are configured in four quadrants, or subassemblies, 

which are separated by a large region of non-voiding water moderator, often referred to as a water cross. 

Tables 17 and 18 contain the design data for the SVEA-64 and SVEA-100 assemblies, respectively. 

The SVEA-64 assembly contains one water rod and 63 fuel rods. The fuel cladding material was not 

specified and is assumed to be Zircaloy. The design dimensions for all SVEA-64 assemblies were 
20 

effectively the same. However, Forsmark 2 assemblies had an Inconel spacer mass given as 35 grams, 

while the Oskarshamn 2 had a spacer mass of 31 grams. Because of the relatively small contribution of 

the spacer material to decay heat, the spacer mass listed in Table 17 for the SVEA-64 assembly is 

35 grams. 

There are no water rods present in any of the SVEA-100 assemblies included in this report (all contain 

100 fuel rods). Fuel clad material was not specified and was assumed to be Zircaloy. 
21 

Table 15 8 × 8 BWR assembly design data 

Design data Type 

1 

Type 

2 

Type 

3 

Type 

4 

Assembly size 8 × 8 8 × 8 8 × 8 8 × 8 

Number of fuel rods 63 63 63 62 

Number of corner rods 12 12 12 0 

Number of water rods 1 1 0 1 

Number of Zr spacer rods 0 0 1 1 

Rod pitch, inner rods (cm) 1.63 1.63 1.63 1.625 

Rod pitch, corner rods (cm) 1.58 1.58 1.58 N/A 

Rod pitch, corner-inner rods (cm) 1.605 1.605 1.605 N/A 

Normal rod diameter (cm) 1.225 1.225 1.225 1.23 

Normal rod clad thickness (cm) 0.074 0.08 0.08 0.082 

Normal rod pellet diameter (cm) 1.058 1.044 1.044 1.044 

Corner rod diameter (cm) 1.175 1.175 1.175 N/A 

Corner rod clad thickness (cm) 0.074 0.08 0.08 N/A 

Corner rod pellet diameter (cm) 1.008 0.994 0.994 N/A 

Fuel clad materiala Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

Water rod outer diameter (cm) 1.225a 1.225a 1.225 1.5 

Water rod thickness (cm) 0.08a 0.08a 0.08a 0.08 

Water rod material Zircaloy-2 Zircaloy-2 Zircaloy-2 Zircaloy-2

Box outer flat-to-flat distance (cm) 13.9 13.9 13.9 13.9 

Box inner flat-to-flat distance (cm) 13.5 13.44 13.44 13.44 

Box wall thickness (cm) 0.2 0.23 0.23 0.23 

Box material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4

Active fuel length (cm) 365.0 368.0 371.2 368.0 

Number of spacers 6 6 6 6 

Spacer material Inconel Inconel Inconel Inconel 

Spacer mass (grams per spacer) 135 135 135 323 

 

a

 Assumed data. 
22 

Table 16 9 × 9 BWR assembly design data 

Design data Type 

1 

Assembly size 9 × 9 

Number of fuel rods 76 

Number of water rods 4 

Number of Zr spacer rods 1 

Fuel rod pitch (cm) 1.445 

Fuel rod diameter (cm) 1.1 

Fuel clad thickness (cm) 0.0665 

Fuel pellet diameter (cm) 0.95 

Fuel clad material Zircaloy-4 a

Water rod outer diameter (cm) 1.5 

Water rod thickness (cm) 0.08 

Water rod material Zircaloy-2 

Box outer flat-to-flat distance (cm) 13.9 

Box wall thickness (cm) 0.23 

Box material Zr-4 

Active fuel length (cm) 368.0 

Number of spacers 6 

Spacer material Zircaloy-4 

Spacer mass (grams per spacer) 323 

 

a

 Assumed data.
23 

Table 17 SVEA-64 assembly design data 

Design data Type 

1 

Assembly size 8 × 8 

Number of fuel rods 63 

Number of water rods 1 

Fuel rod pitch (cm) 1.58 

Fuel rod diameter (cm) 1.225 

Fuel clad thickness (cm) 0.08 

Fuel pellet diameter (cm) 1.044 

Water rod outer diameter (cm) 1.225 

Water rod thickness (cm) 0.08 

Box outer flat-to-flat distance (cm) 14.0 

Box wall thickness (cm) 0.11 

Box material Zircaloy-4 

Water cross material Zircaloy-4 

Water cross wall thickness (cm) 0.08 

Sub-assembly inner measureb (cm) 6.59 

Active fuel length (cm) 368.0 

Fuel clad material Zircaloy-4 a

Water rod material Zircaloy-2 

Number of spacers 24 (4 × 6) 

Spacer material Inconel 

Spacer mass (grams per spacer) 35 

 

a

 Assumed data. 

b

 Flat-to-flat distance across the subassembly box.
24 

Table 18 SVEA-100 assembly design data 

Design data Type 

1 

Assembly size 10 × 10 

Number of fuel rods 100 

Number of water rods 0 

Fuel rod pitch (cm) 1.24 

Fuel rod diameter (cm) 0.962 

Fuel clad thickness (cm) 0.063 

Fuel pellet diameter (cm) 0.819 

Fuel clad material Zircaloy-4 a

Box outer flat-to-flat distance (cm) 13.96 

Box wall thickness (cm) 0.11 

Box material Zircaloy-4 

Water cross material Zircaloy-4 

Water cross wall thickness (cm) 0.11 

Sub-assembly inner dimensionb (cm) 6.56 

Nominal active fuel length (cm) 370.0 

Number of spacers 24 (4 × 6) 

Spacer material Inconel 

Spacer mass (grams per spacer) 24 

 

a

 Assumed data. 

b

 Flat-to-flat distance across the subassembly box.
25 

5 FUEL DATA 

This section describes the enrichment levels, fuel density, and burnable poison rod data for the fuel 

assemblies measured and included in the present report. For most assemblies only 235U and 238U 

concentrations were available and are listed. For the 15 × 15 assemblies, data on 234U and 236U were also 

available. 

The fuel enrichment in PWR assemblies is typically the same for all non-poison rods in the assembly. 

However, BWR assemblies typically reduce the enrichment of the outermost rods of the assembly (where 

thermal flux is larger) relative to the rods internal to the assembly to achieve a more uniform power 

distribution. Assembly enrichment zoning patterns vary widely and are typically proprietary fuel vendor 

information. The assembly enrichment zoning was not provided for the Swedish assemblies. All 

enrichment data provided in the tables of this section represent assembly-average values. The enrichment 

zoning patterns shown for the BWR assemblies in Appendix A.2 is for illustrative purposes only. 

5.1 PWR 15 × 15 Assemblies

Table 19 shows fuel data for the 15 × 15 assemblies. Reactor, assembly, and assembly type are noted. 

The complete design and fuel data for each assembly type can be obtained by cross-referencing the type 

with the design data given in Section 4. The effective fuel (UO2) density accounts for porosity and 

dishing of the fuel pellets. The initial composition of the uranium is listed by the isotopic vector for 234U, 

235U, 236U, and 238U, and also by mass in the assembly. There was no burnable absorber rod exposure 

indicated for any of the measured 15 × 15 assemblies. 

5.2 PWR 17 × 17 Assemblies

Table 20 shows fuel data for the 17 × 17 assemblies. Reactor, assembly, and assembly type are noted. 

The effective fuel density and uranium isotopic mass values are listed for 235U and 238U. The initial 

concentrations of 234U and 236U were not provided. 

Four of the measured 17 × 17 assemblies were identified as having burnable absorber rods during 

irradiation. However, the duration of the BA rod insertion was not provided. Based on typical PWR 

operational practice,13 the BA rods are typically inserted in the assemblies during the first cycle of reactor 

exposure, after which they are withdrawn. Although the boron levels in the poison rods are significantly 

depleted after the first cycle, the BA rods are not always withdrawn after one cycle, and continued 

presence of the rods will reduce the moderation in the guide tubes. 

5.3 BWR 8 × 8 Assemblies

Table 21 shows fuel data for the 8 × 8 assemblies. Isotopic data for uranium are listed for the 235U and 

238U isotopes. The number of integral burnable poison rods and the concentration of burnable poison 

(Gd2O3) in the fuel pellets that contain poison are also listed. The burnable poison rods are an integral 

part of the assembly, unlike the BA rods used in the PWR assemblies that can be inserted or withdrawn 

from the guide tubes of the assembly. The table also lists the relative poison, a term that refers to the 

fraction of fuel pellets in the fuel rod that contain Gd2O3. Therefore, a value of 1.0 indicates that all of the 

pellets in the poison rods contain gadolinium. 
26 

5.4 BWR 9 × 9 Assemblies

The 9 × 9 assembly data are listed in Table 22. These data are of similar type to those for the 8 × 8 

assemblies. 

5.5 SVEA Assemblies

Data for SVEA-64 and SVEA-100 assemblies are listed in Tables 23 and 24, respectively. For the 

SVEA-64 assemblies, only the 235U and total uranium concentrations were provided. The initial 

concentrations for 234U and 236U may be estimated using the uranium isotopic vectors provided for the 

15 × 15 assemblies in Table 19 by assuming that 234U and 236U scale proportionately with the initial 235U 

concentration. 
27 

Table 19 Fuel data for PWR 15 × 15 assemblies 

Reactor Assembly ID Type Fuel density (g/cm3) Eff. fuel density (g/cm3) Initial U (g) Initial 238U (g) Initial 235U (g) Initial 236U (g) Initial 234U (g) Initial 238U (wt %) Initial 235U (wt %) Initia

236U (wt %

Ringhals 2 C01 1 10.41 10.23 455,789 441,485 14,107 69 128 96.862 3.095 0.015 Ringhals 2 C12 1 10.41 10.18 453,736 439,496 14,043 69 128 96.862 3.095 0.015 Ringhals 2 C20 1 10.41 10.21 454,758 440,486 14,075 69 128 96.862 3.095 0.015 Ringhals 2 D27 2 10.41 10.10 432,589 418,326 14,066 69 128 96.703 3.252 0.016 Ringhals 2 D38 2 10.41 10.13 434,214 419,898 14,119 69 128 96.703 3.252 0.016 Ringhals 2 E38 2 10.41 10.12 433,593 419,527 13,869 69 128 96.756 3.199 0.016 Ringhals 2 E40 2 10.41 10.13 434,244 420,157 13,890 69 128 96.756 3.199 0.016 Ringhals 2 F14 2 10.41 10.18 436,382 422,232 13,953 69 128 96.757 3.197 0.016 Ringhals 2 F21 2 10.41 10.17 435,939 421,803 13,939 69 128 96.757 3.197 0.016 Ringhals 2 F25 2 10.41 10.21 437,286 423,107 13,982 69 128 96.757 3.197 0.016 Ringhals 2 F32 2 10.41 10.20 436,993 422,824 13,972 69 128 96.758 3.197 0.016 Ringhals 2 G11 2 10.41 10.18 436,180 422,077 13,906 69 128 96.767 3.188 0.016 Ringhals 2 G23 2 10.41 10.18 436,125 421,947 13,981 69 128 96.749 3.206 0.016 Ringhals 2 I09 2 10.41 10.21 437,353 423,149 14,007 69 128 96.752 3.203 0.016 Ringhals 2 I24 2 10.41 10.03 429,597 415,641 13,759 69 128 96.751 3.203 0.016 Ringhals 2 I25 2 10.41 10.11 433,062 418,995 13,870 69 128 96.752 3.203 0.016 
28 

Table 20 Fuel data for PWR 17 × 17 assemblies 

Reactor Assembly ID Type Fuel density (g/cm3) Eff. fuel density (g/cm3) Initial U (g) Initial 238U (g) Initial 235U (g) Initial 238U (wt %) 

Ringhals 3 2A5 1 10.45 10.31 462,026 452,284 9,742 97.900 Ringhals 3 5A3 1 10.45 10.3 461,477 451,787 9,690 97.900 Ringhals 3 0C9 1 10.45 10.21 457,639 443,448 14,191 96.899 Ringhals 3 1C2 1 10.45 10.24 459,050 444,815 14,235 96.899 Ringhals 3 1C5 1 10.45 10.22 457,992 443,790 14,202 96.899 Ringhals 3 2C2 1 10.45 10.25 459,490 445,241 14,249 96.899 Ringhals 3 3C1 1 10.45 10.23 458,433 444,217 14,216 96.9 Ringhals 3 3C5 1 10.45 10.24 458,873 444,643 14,230 96.899 Ringhals 3 3C9 1 10.45 10.25 459,138 444,900 14,238 96.899 Ringhals 3 4C4 1 10.45 10.24 459,050 444,815 14,235 96.899 Ringhals 3 4C7 1 10.45 10.23 458,256 444,045 14,211 96.899 Ringhals 3 0E2 1 10.45 10.35 463,598 449,111 14,487 96.897 Ringhals 3 0E6 1 10.45 10.31 461,769 447,453 14,316 96.897 Ringhals 3 1E5 1 10.45 10.35 463,898 449,506 14,392 96.897 
29 

Table 21 Fuel data for BWR 8 × 8 assemblies 

Reactor Assembly ID Type Fuel density (g/cm3) Eff. fuel density (g/cm3) Initial U (g) Initial 238U (g) Initial 235U (g) Initial 235U (wt %) Number of poison rods 

Ringhals 1 1177 1 10.26 10.16 180,587 175,816 4,771 2.642 3 Ringhals 1 1186 1 10.43 10.32 180,515 175,750 4,765 2.640 3 Ringhals 1 6423 2 10.46 10.35 177,701 172,547 5,154 2.900 4 Ringhals 1 6432 2 10.45 10.34 177,520 172,382 5,138 2.894 4 Ringhals 1 6454 2 10.46 10.35 177,683 172,534 5,149 2.898 4 Ringhals 1 8327 2 10.45 10.34 177,544 172,389 5,155 2.904 4 Ringhals 1 8331 2 10.46 10.35 177,690 172,520 5,170 2.910 4 Ringhals 1 8332 2 10.45 10.34 177,519 172,380 5,139 2.895 4 Ringhals 1 8338 2 10.45 10.35 177,596 172,426 5,170 2.911 4 Oskarshamn 3 12078 2 10.44 10.33 177,358 172,787 4,571 2.577 3 Forsmark 2 5535 2 10.50 10.35 177,689 173,966 3,723 2.095 0 Forsmark 1 3838 2 10.50 10.36 177,903 174,192 3,711 2.090 3 Barsebäck 1 9329 3 10.50 10.32 178,771 173,548 5,223 2.922 5 Barsebäck 1 10288 3 10.50 10.34 179,159 173,868 5,291 2.953 5 Barsebäck 2 14076 3 10.50 10.37 179,571 173,906 5,665 3.154 4 Oskarshamn 2 1377 3 10.71 10.60 183,575 179,535 4,040 2.201 0 Oskarshamn 2 1389 3 10.72 10.61 183,650 179,608 4,042 2.201 0 Oskarshamn 2 1546 3 10.73 10.63 183,968 179,919 4,049 2.201 0 Oskarshamn 2 1696 3 10.75 10.64 184,253 180,198 4,055 2.201 0 Oskarshamn 2 1704 3 10.74 10.63 184,022 179,972 4,050 2.201 0 Oskarshamn 2 2995 3 10.47 10.36 179,382 174,540 4,842 2.699 4 Oskarshamn 2 6350 3 10.44 10.34 179,003 173,857 5,146 2.875 6 Forsmark 1 KU0100 4 10.30 10.17 174,920 169,730 5,190 2.970 4 
30 

Table 22 Fuel data for BWR 9 × 9 assemblies 

Reactor Assembly 

ID Type 

Fuel 

density 

(g/cm3

) 

Eff. fuel 

density 

(g/cm3) 

Initial 

U 

(g) 

Initial

235U 

(g) 

Initial 

235U 

(wt %)

Number 

of 

poison 

rods 

Gd2O3

(wt %)

Relative 

poison 

Forsmark 1 KU0269 1 10.30 10.14 177,020 5,201 2.938 6 2.5 1 

Forsmark 1 KU0278 1 10.30 10.15 177,133 5,206 2.938 6 2.5 1 

Forsmark 1 KU0282 1 10.30 10.15 177,097 5,204 2.938 6 2.5 1 
31 

Table 23 Fuel data for SVEA-64 assemblies 

Reactor Assembly ID Type Fuel density (g/cm3) Eff. fuel density (g/cm3) Initial U (g) Initial 235U (g) Initial 235U (wt %) Number of poison rods Gd2O3 (wt %

Forsmark 2 11494 1 10.50 10.36 181,088 5,287 2.920 4 2.55 Forsmark 2 11495 1 10.50 10.36 181,070 5,287 2.910 4 2.55 Forsmark 2 13775 1 10.50 10.38 181,340 5,175 2.850 5 2.55 Oskarshamn 2 12684 1 10.45 10.34 182,320 5,291 2.902 6 3.15 

Table 24 Fuel data for SVEA-100 assemblies 

Reactor Assembly ID Type Fuel density (g/cm3) Eff. fuel density (g/cm3) Initial U (g) Initial 238U (g) Initial 235U (g) Initial 235U (wt %) Number of poison rods Gd

(wt 

Oskarshamn 3 13628 1 10.50 10.39 180,774 175,873 4,901 2.711 8 4Oskarshamn 3 13630 1 10.50 10.39 180,775 175,874 4,901 2.711 8 4Forsmark 3 13847 1 10.70 10.58 180,669 175,667a 5,002 2.770 5 2.Forsmark 3 13848 1 10.70 10.58 180,667 175,665a 5,002 2.770 5 2.

a Values derived from 235U values. Note that the derived values exclude contributions from 234U and 236U.
32 
33 

6 POWER HISTORY DATA 

6.1 Assembly Burnup Data

The power history data for the measured assemblies are detailed in Tables 25–34. There is one table for 

each reactor. For each assembly, the assembly design and the assembly identifier are listed. The 

exposure history information is listed as burnup per cycle together with cycle length in days. The total 

burnup for the assembly, obtained from the sum of cycle burnups for each cycle exposed, is also provided 

in the tables. The down time is the time between consecutive cycles, generally for the purposes of 

refueling. The average assembly specific power (MW/MTU) for a cycle can be derived from the cycle 

burnup (MWd/MTU) divided by the cycle length (days). The cycle-average power on an assembly basis 

(MW/assembly) can be obtained using the assembly initial uranium mass values listed in Section 5. Note 

that some cycles are subdivided to account for significant periods of low and zero operating power during 

a cycle. 

Table 25 Barsebäck 1 assembly irradiation history 

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 Design Assembly Burnup (MWd/MTU) 

9329 8 × 8 8,219 17,727 25,034 31,958 41,127 

10288 8 × 8 9,726 17,440 25,587 35,218 

Table 26 Barsebäck 2 assembly irradiation history 

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 Design Assembly Burnup (MWd/MTU) 

14076 8 × 8 8,398 18,616 24,130 34,126 40,010 
Table 27 Forsmark 1 assembly irradiation history 

Operational data type Cycle 1A Cycle 1B Cycle 2 Cycle 3 Cycle 4 Cycle 5 Cycle 6 Cycle 7 Cycle 8 CyStartup 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 8/18Shutdown 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 8/17Operating days 428 307 324 343 298 361 374 296 369 Downtime days 52 67 21 24 38 18 19 30 35 

Assembly ID Design Assembly Burnup (MWd/MTU) 

3838 8 × 8 6,160 13,092 15,990 19,053 21KU0100 8 × 8 10,643 18,706 26,630 34,193 KU0269 9 × 9 10,158 18,787 27,153 35,113 KU0278 9 × 9 10,414 17,685 24,764 29,413 35KU0282 9 × 9 10,122 18,800 25,844 32,112 37

Table 28 Forsmark 2 assembly irradiation history 

Operational data type Cycle 1A Cycle 1B Cycle 2 Cycle 3 Cycle 4 Cycle 5 Cycle 6 Cycle 7 Cycl

Startup date 12/1/1980 8/18/1982 6/11/1983 7/13/1984 7/26/1985 8/9/1986 6/20/1987 8/4/1988 9/4/1Shutdown date 5/1/1982 5/20/1983 5/25/1984 7/5/1985 7/25/1986 5/31/1987 7/15/1988 8/18/1989 5/18/1Operating days 516 275 349 357 364 295 391 379 Downtime days 109 22 49 21 15 20 20 17 

Assembly ID Design Assembly Burnup (MWd/MTU) 

5535 8 × 8 8,850 10,758 12,862 14,982 16,665 18,115 19,944 

11494 SVEA-64 9,098 16,676 24,136 32,431 11495 SVEA-64 9,098 16,676 24,136 32,431 13775 SVEA-64 8,801 18,300 23,975 27,4

34 
Table 29 Forsmark 3 assembly irradiation history 

Operational data type 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 ID Design Assembly Burnup (MWd/MTU) 

13847 SVEA-100 7,857 17,217 23,779 31,275 13848 SVEA-100 7,857 17,217 23,779 31,275 

Table 30 Oskarshamn 2 assembly irradiation history 

Operational data type Cycle 1 Cycle 2 Cycle 3 Cycle 4 Cycle 5 Cycle 6 Cycle 7 Cycle 8 CyclStartup 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 9/19/1Shutdown 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 7/1/1Operating days 562 315 321 320 317 329 327 361 Downtime days 77 86 71 41 39 46 31 31 

Assembly ID Design Assembly Burnup (MWd/MTU) 

1377 8 × 8 8,982 14,546 1389 8 × 8 4,315 7,300 11,829 14,539 16,933 19,481 1546 8 × 8 6,961 13,840 21,032 24,470 1696 8 × 8 4,492 7,904 13,258 16,319 18,502 20,870 1704 8 × 8 4,587 8,031 13,384 16,638 19,437 2995 8 × 8 6,143 13,359 19,705 23,577 29,978 6350 8 × 8 7,113 14,909 24,019 

35 
Table 30 Oskarshamn 2 assembly irradiation history (cont.) 

Operational data type Cycle 11 Cycle 12 Cycle 13 Cycle 14 Cycle 15 Cycle 16 Startup date 7/13/1985 9/27/1986 9/4/1987 9/10/1988 9/1/1989 10/11/1990 Shutdown date 8/16/1986 7/31/1987 8/20/1988 8/5/1989 8/10/1990 8/2/1991 Operating days 399 307 351 329 343 295 Downtime days 42 35 21 27 62 --- 

Assembly ID Design Assembly Burnup (MWd/MTU) 

12684 SVEA-64 11,134 20,032 28,184 35,495 43,084 46,648 

Table 31 Oskarshamn 3 assembly irradiation history 

Operational data type Cycle 1 Cycle 2 Cycle 3 Cycle 4 Cycle 4B Cycle 5 Cycle 6 Startup date 3/18/1985 7/26/1986 7/26/1987 8/14/1988 12/22/1988 6/20/1989 8/1/1990 Shutdown date 7/4/1986 7/3/1987 7/8/1988 12/10/1988 6/7/1989 6/23/1990 6/24/199Operating days 473 342 348 118 167 368 327 Downtime days 22 23 37 12 13 39 --- 

Assembly ID Design Assembly Burnup (MWd/MTU) 

12078 8 × 8 8,690 16,948 25,160 

13628 SVEA-100 9,171 18,613 20,910 24,673 32,885 35,619 13630 SVEA-100 9,328 18,984 21,970 25,793 32,866 40,363 

36 
Table 32 Ringhals 1 assembly irradiation history 

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 

37 
Table 33 Ringhals 2 assembly irradiation history 

Operational data type Cycle 1 Cycle 2 Cycle 3 Cycle 4 Cycle 5 Cycle 6 Cycle 7 Cycle 8 Cycle 9 Cycle 1

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 6/14/198Shutdown 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 4/30/198Operating days 1029 267 312 281 290 317 273 260 266 32Downtime days 85 56 83 78 80 84 91 90 71 6

Assembly 

ID 

Design Assembly Burnup (MWd/MTU) 

C01 15 × 15 11,247 20,650 28,219 36,688 C12 15 × 15 11,247 20,565 27,955 36,385 C20 15 × 15 11,247 20,624 28,078 35,720

D27 15 × 15 9,510 22,399 31,666 39,676 D38 15 × 15 6,367 15,698 23,056 31,757 39,403 

E38 15 × 15 7,568 16,026 25,905 33,973 E40 15 × 15 7,705 14,954 25,609 34,339 

F14 15 × 15 5,069 15,824 25,722 34,009 F21 15 × 15 4,767 11,084 21,130 29,385 36,273 F25 15 × 15 8,307 19,056 27,372 35,352 F32 15 × 15 10,553 21,162 29,553 37,314 

G11 15 × 15 6,890 17,312 25,180 32,123 35,463 G23 15 × 15 10,268 20,303 27,921 35,633 

I09 15 × 15 6,727 15,677 24,742 I24 15 × 15 8,245 17,212 26,356 34,294 I25 15 × 15 5,207 10,198 20,001 28,999 

38 
Table 34 Ringhals 3 assembly irradiation history 

Operational data type Cycle 1A Cycle 1B Cycle 2 Cycle 3 Cycle 4 CyclStartup date 7/29/1980 9/14/1983 7/24/1984 7/11/1985 7/18/1986 8/4/19Shutdown date 6/2/1983 5/11/1984 5/25/1985 5/30/1986 6/18/1987 7/7/19Operating days 1038 240 305 323 335 3Downtime days 104 74 47 49 47 Assembly ID Design Assembly Burnup (MWd/MTU) 

2A5 17 × 17 12,228 20,107 5A3 17 × 17 11,696 19,699 

0C9 17 × 17 9,884 18,076 28,426 38,442 1C2 17 × 17 6,249 11,268 22,777 33,318 1C5 17 × 17 9,884 17,986 28,397 38,484 2C2 17 × 17 7,783 16,128 26,060 36,577 3C1 17 × 17 7,783 16,124 26,055 36,572 3C5 17 × 17 9,884 17,997 28,340 38,373 3C9 17 × 17 7,783 16,160 26,036 36,560 4C4 17 × 17 6,249 11,240 22,270 33,333 4C7 17 × 17 9,884 17,985 28,332 38,370 

0E2 17 × 17 7,496 20,530 31,838 41,60E6 17 × 17 12,490 25,521 35,91E5 17 × 17 10,556 23,690 34,6

39 
40 

6.2 Assembly Burnup Profile Data

Assembly axial burnup profiles were available for some but not all of the assemblies. The profiles are 

described in terms of the burnup in discrete axial zones of the assembly. Each zone has the same axial 

height. In the case of PWR assemblies, there were 24 axial zones, and for BWR assemblies there were 

25 zones. The local zone burnup data are given in units of megawatt days per metric ton of uranium 

(MWd/MTU) for each axial zone. The burnup values are those reported by the utilities. Note that values 

for the Barsebäck and Oskarshamn assemblies are reported to the nearest 100 MWd/MTU. 

Axial burnup profile data were available for all measured PWR assemblies. All of the 15 × 15 assemblies 

came from the Ringhals 2 reactor, and all of the 17 × 17 assemblies were from the Ringhals 3 reactor. 

The corresponding axial profiles are listed in Tables 35 and 36, respectively. The axial burnup profiles 

for the 8 × 8, 9 × 9, SVEA-64, and SVEA-100 assemblies are listed in Tables 37–40. Those assemblies 

for which profiles were not available are not listed in the tables. The only assemblies for which axial 

profile data were not available were Ringhals-1 8 × 8 assemblies 6423, 6432, 6454, 8327, 8331, 8332, 

and 8338. 

If burnup profiles are desired for assemblies for which data are not available, it may be possible to derive 

representative profiles by evaluating similar assemblies that have listed profiles. Following the axial 

burnup profile tables, there is an evaluation and discussion of the profile data. 
Table 35 PWR 15 × 15 assembly burnup profile data 

Reactor Ringhals 2

Assembly ID C01 C12 C20 D27 D38 E38 E40 

Axial level Discharge burnup (MWd/MTU) 

1 (bottom) 20938 20894 20317 22790 22420 17154 20163 2 30055 29982 29205 33099 32169 26466 28758 3 35077 34977 34139 38570 37501 31559 33332 4 37367 37245 36428 40893 39961 34119 35324 5 38321 38185 37414 41811 41053 35407 36117 6 38677 38530 37799 42151 41538 36139 36402 7 38786 38631 37930 42290 41767 36518 36488 8 38808 38647 37966 42307 41887 36706 36505 9 38807 38641 37975 42295 41961 36829 36503 

10 38808 38636 37983 42286 42020 36929 36501 11 38815 38640 37996 42284 42076 37019 36502 12 38829 38651 38016 42290 42132 37110 36512 13 38850 38669 38042 42303 42187 37202 36529 14 38877 38694 38074 42309 42239 37292 36541 15 38912 38726 38111 42317 42288 37379 36555 16 38957 38769 38158 42331 42336 37462 36574 17 39015 38824 38215 42357 42379 37535 36598 18 39077 38884 38272 42390 42403 37577 36622 19 39105 38910 38288 42400 42364 37537 36618 20 38981 38785 38147 42279 42142 37286 36495 21 38400 38206 37548 41722 41430 36525 35994 22 36639 36455 35788 39964 39490 34576 34452 23 32133 31973 31350 35286 34684 30023 30358 

24 (top) 23157 23042 22574 25664 25094 21354 21966 

41 
Table 35 PWR 15 × 15 assembly burnup profile data (continued)

Reactor Ringhals 2 

Assembly ID F14 F21 F25 F32 G11 G23 I09 I24 

Axial level Discharge burnup (MWd/MTU) 

1 (bottom) 19083 21231 19894 32227 19843 21047 24212 19588 2

2 27514 30092 28531 44459 28645 29748 34163 28041 33 32196 34781 33299 50497 33541 34367 39284 32689 34 34384 36861 35525 52964 35865 36409 41482 34872 35 35380 37765 36544 53879 36946 37275 42380 35886 36 35845 38151 37020 54165 37455 37635 42703 36356 37 36065 38306 37253 54214 37706 37782 42790 36576 38 36174 38368 37378 54181 37840 37842 42793 36684 39 36238 38394 37457 54125 37925 37870 42771 36745 3

10 36287 38407 37518 54067 37988 37887 42748 36787 311 36330 38416 37572 54012 38044 37903 42729 36822 312 36373 38424 37624 53963 38098 37920 42715 36855 313 36416 38432 37675 53917 38152 37936 42705 36889 314 36461 38440 37726 53876 38206 37954 42699 36922 315 36506 38450 37775 53839 38261 37974 42697 36956 316 36551 38459 37821 53808 38314 37996 42699 36987 317 36592 38466 37859 53781 38359 38017 42703 37011 318 36616 38459 37871 53749 38376 38028 42700 37010 319 36584 38406 37817 53676 38319 37997 42658 36944 320 36391 38212 37587 53452 38071 37832 42478 36710 321 35772 37619 36916 52759 37349 37279 41888 36046 322 34084 35970 35152 50752 35486 35690 40178 34320 323 29887 31731 30832 45373 31015 31546 35655 30099 3

24 (top) 21550 23080 22276 33752 22295 23035 26135 21747 2

42 
Table 36 PWR 17 × 17 assembly burnup profile data 

Reactor Ringhals 3 

Assembly ID 2A5 5A3 0C9 1C2 1C5 2C2 3C1 3C5 

Axial level Discharge burnup (MWd/MTU) 

1 (bottom) 9653 8661 21140 18077 21160 20091 20090 21119 

2 14880 14073 30635 26607 30666 29186 29185 30603 3 18290 17624 36170 31538 36208 34451 34449 36127 4 20213 19664 38931 33882 38970 37025 37022 38877 5 21233 20897 40254 34933 40291 38210 38206 40193 6 21734 21421 40852 35372 40886 38723 38719 40784 7 21953 21617 41103 35579 41136 38939 38935 41031 8 22029 21669 41204 35656 41236 39023 39019 41129 9 22038 21660 41248 35691 41280 39055 39051 41171 

10 22025 21633 41275 35719 41307 39077 39072 41197 11 22012 21612 41302 35749 41335 39107 39102 41222 12 22010 21605 41334 35781 41368 39148 39143 41254 13 22023 21616 41373 35817 41409 39202 39197 41292 14 22052 21645 41419 35855 41457 39269 39263 41338 15 22096 21690 41472 35897 41512 39348 39342 41390 16 22150 21748 41530 35942 41573 39440 39433 41448 17 22200 21805 41585 35988 41630 39541 39533 41502 18 22219 21836 41612 36019 41662 39630 39622 41529 19 22145 21781 41551 35990 41605 39654 39644 41467 20 21854 21520 41251 35778 41311 39468 39458 41168 21 21112 20821 40373 35088 40440 38741 38730 40293 22 19488 19256 38178 33252 38247 36739 36728 38103 23 16251 16092 33136 28872 33201 31955 31946 33073 

24 (top) 10915 10834 23686 20553 23733 22837 22831 23640 

43 
Table 36 PWR 17 × 17 assembly burnup profile data (continued) 

Reactor Ringhals 3 

Assembly ID 4C4 4C7 0E2 0E6 1E5 

Axial level Discharge burnup (MWd/MTU) 

1 (bottom) 18075 21118 24593 20541 18931 2 26607 30601 34686 29342 27564 3 31541 36124 40170 34267 32452 4 33888 38874 42652 36574 34844 5 34942 40189 43723 37597 36023 6 35383 40780 44169 38045 36563 7 35591 41027 44347 38248 36844 8 35670 41125 44413 38347 36985 9 35706 41167 44435 38405 37072 

10 35735 41193 44438 38446 37136 11 35765 41219 44435 38482 37193 12 35798 41250 44428 38516 37246 13 35834 41289 44419 38550 37298 14 35873 41335 44410 38584 37350 15 35915 41387 44400 38619 37403 16 35961 41445 44391 38657 37454 17 36007 41499 44381 38696 37499 18 36039 41527 44361 38729 37522 19 36011 41466 44294 38712 37478 20 35800 41167 44065 38507 37257 21 35111 40292 43366 37827 36565 22 33273 38103 41416 35990 34713 23 28892 33073 36490 31481 30256 

24 (top) 20567 23641 26583 22679 21673 

44 
Table 37 BWR 8 × 8 assembly burnup profile data 

Reactor Ringhals 1 Ringhals 1 Oskarshamn 3 Forsmark 1 Forsmark 2 Barsebäck 1 Barsebäck 1 

Assembly ID 1177 1186 12078 3838 5535 9329 10288 

Axial level Discharge burnup (MWd/MTU) 

1 (bottom) 20521 17481 11300 10899 9028 30200 18300 2 30692 26029 19500 19258 15826 31900 29600 3 35545 30230 23600 23070 18445 37700 35100 4 37745 32082 25400 24864 19467 40300 37300 5 39059 33061 27000 26611 20513 42600 39200 6 39836 33591 27700 27703 21028 43900 40000 7 40202 33899 27900 28282 21157 44700 40200 8 40412 34072 27700 28538 20993 44700 39700 9 40658 34225 28300 29705 21543 45800 40300 

10 40701 34303 28600 30180 21826 46200 40400 11 40725 34310 28600 30167 21923 46500 40300 12 40538 34151 28300 29904 21944 46200 39600 13 40498 34141 28900 30323 22558 46700 39900 14 40380 34057 29300 30356 22882 46800 37700 15 40276 33908 29200 30116 22953 46500 39200 16 39922 33607 28700 29719 22965 45800 38200 17 39672 33376 28900 29633 23121 46000 38100 18 39220 32891 28600 29031 22802 45700 37600 19 38483 32336 27900 27866 21956 44900 36700 20 37499 31411 26600 26758 21263 43500 35100 21 36039 30197 25700 25237 20175 42400 33900 22 33892 28417 23900 22794 18347 39900 31600 23 30267 25316 21000 19043 15429 35700 28000 24 24808 20626 16700 14381 11730 22900 23300 

25 (top) 17579 14378 9400 10623 8835 22400 17600 

45 
Table 37 BWR 8 × 8 assembly burnup profile data (continued) 

Reactor Oskarshamn 2 Oskarsham 2 Oskarshamn 2 Oskarshamn 2 Oskarshamn 2 Oskarshamn 2 Oskarshamn 2 

Assembly ID 1377 1389 1546 1696 1704 2995 6350 

Axial level Discharge burnup (MWd/MTU) 

1 (bottom) 7600 11000 13000 12100 11100 19500 16300 2 12100 16700 19800 18400 17000 27900 24700 3 14800 20000 23600 21700 20300 32100 29000 4 16300 21700 25600 23400 22100 34100 30800 5 17400 22700 26800 24200 23000 34900 31600 6 17600 23200 27300 24600 23500 35100 31800 7 17400 23400 27400 24800 23600 35000 31800 8 17000 23500 27600 24800 23500 34900 31600 9 16800 23400 27600 24700 23300 34600 31600 

10 16600 23300 27500 24600 23100 34200 31400 11 16400 23000 27600 24400 22800 33900 31500 12 16300 22700 28000 24100 22400 33400 31200 13 16200 22400 28200 23800 22100 33000 31000 14 16100 22100 28100 23400 21700 32600 30700 15 16000 21700 28000 23100 21300 32200 30300 16 15800 21300 27900 22700 21000 31900 29800 17 15600 20800 27800 22300 20600 31400 29500 18 15300 20300 27400 21800 20100 30900 29000 19 15000 19600 26600 21100 19400 30200 28400 20 14500 18700 25500 20300 18700 29400 27400 21 13800 17600 24000 19100 17500 28100 26200 22 12800 16100 22000 17500 16000 26100 24400 23 11300 13900 19100 15200 13800 23100 21700 24 8900 10900 15100 12000 10800 18500 17600 

25 (top) 5900 6900 10200 7700 7000 12600 12500 

46 
47 

Table 38 BWR 9 × 9 assembly burnup profile data 

Reactor Forsmark 1 

Assembly ID KU0269 KU0278 KU0282 

Axial level Discharge burnup (MWd/MTU) 

1 (bottom) 8603 7943 9352 

2 28322 25901 29564 

3 35731 32745 37022 

4 38495 35584 39994 

5 40365 37733 42134 

6 40905 38658 42902 

7 40872 39039 43014 

8 40206 38705 42388 

9 40887 39496 42954 

10 41093 39734 43000 

11 41171 39840 43023 

12 40924 39567 42633 

13 41606 40261 43280 

14 41500 40282 43214 

15 40976 40025 42853 

16 40235 39637 42287 

17 40117 40025 42500 

18 39361 39897 42154 

19 37909 39247 41216 

20 36592 38879 40570 

21 34472 37657 39121 

22 31013 34810 36152 

23 25572 29326 30726 

24 16313 18760 19973 

25 (top) 8830 10314 11201 
48 

Table 39 SVEA-64 assembly burnup profile data 

Reactor Oskarshamn 2 Forsmark 2 Forsmark 2 Forsmark 2 

Assembly ID 12684 11494 11495 13775 

Axial level Discharge burnup (MWd/MTU) 

1 (bottom) 23300 16962 16962 16247 

2 37600 27585 27585 26650 

3 44300 32280 32280 31970 

4 47000 33996 33996 34198 

5 49400 35386 35386 36158 

6 50700 36003 36003 37084 

7 51300 36061 36061 37443 

8 51200 35683 35683 37216 

9 52300 36114 36114 37871 

10 52800 36279 36279 38056 

11 53000 36234 36234 38058 

12 52400 35980 35980 37486 

13 53000 36329 36329 37776 

14 53100 36322 36322 37614 

15 52700 35997 35997 37182 

16 51800 35791 35791 36356 

17 51900 35831 35831 36336 

18 51400 35446 35446 35843 

19 50400 34534 34534 34978 

20 48500 33710 33710 33511 

21 47100 32281 32281 32378 

22 44300 29702 29702 30207 

23 39600 25442 25442 26735 

24 32300 19667 19667 21538 

25 (top) 24700 16170 16170 12047 
49 

Table 40 SVEA-100 assembly burnup profile data 

Reactor Oskarshamn 3 Oskarshamn 3 Forsmark 3 Forsmark 3 

Assembly ID 13628 13630 13847 13848 

Axial level Discharge burnup (MWd/MTU) 

1 (bottom) 7800 9400 9053 9053 

2 27300 31800 26749 26749 

3 34100 39200 32086 32086 

4 37000 42300 34240 34240 

5 39100 44700 35674 35674 

6 40100 45700 36261 36261 

7 40400 46200 36329 36329 

8 40200 46000 36073 36073 

9 40900 46700 36565 36565 

10 41200 47000 36809 36809 

11 41200 46900 36798 36798 

12 40700 46300 36625 36625 

13 41100 46700 36875 36875 

14 41100 46600 36784 36784 

15 40900 46200 36451 36451 

16 40400 45500 36363 36363 

17 40800 45700 36474 36474 

18 40600 45300 36262 36262 

19 40100 44500 35592 35592 

20 38700 42900 34934 34934 

21 37500 41500 33500 33500 

22 34800 38700 30676 30676 

23 30500 34100 25833 25833 

24 24100 27700 19696 19696 

25 (top) 9800 11500 7910 7910 
50 

6.3 Burnup-Profile Trends

Axial burnup profiles for selected PWR assemblies from Ringhals-3 are shown in Figure 4. The plotted 

values are in units of gigawatt days per metric ton uranium (GWd/MTU). The profiles are plotted as a 

function of and fractional assembly height, where a value of 1 corresponds to the top of the fuel assembly. 

As is typical of PWR assemblies, the profiles are relatively symmetric about the assembly center plane. 

The burnup profiles shown here are representative of the other discharged assemblies. 

Figure 4 PWR burnup profiles for Ringhals 3 reactor assemblies 

Burnup profiles were not available for all BWR assemblies. BWR burnup profiles (as compared to PWR 

ones) are generally not symmetric about the assembly centerline height because of the axial variation in 

void, and thus moderator density, and the insertion of control blades. The burnup profile for assembly 

14076 from the Barsebäck 2 BWR reactor is illustrated in Figure 5 (again, the plotted values are 

GWd/MTU). This profile is not symmetric; it peaks in the lower half of the assembly where the 

moderator density is highest. A selection of BWR burnup profiles for assemblies from the Oskarshamn 2 

reactor is shown in Figure 6. Because the profiles represent discharge values, they do not reflect possible 

cycle-to-cycle variations in axial power during irradiation. 

0

10

20

30

40

50

60

0 0.2 0.4 0.6 0.8 1

2A5

5A3

0C9

1C2

2C2

3C1

Burnup (GWd/t)

Fractional height
51 

0

10

20

30

40

50

60

0 0.2 0.4 0.6 0.8 1

Burnup (GWd/t)

Fractional height

Figure 5 BWR burnup profile for Barsebäck 2 assembly 14076 

0

10

20

30

40

50

60

0 0.2 0.4 0.6 0.8 1

1377

1389

1546

1696

1704

2995

3054

3058

3064

6350

12684

Burnup (GWd/t)

Fractional height

Figure 6 BWR burnup profiles for selected Oskarshamn 2 assemblies 

(Note: Several assemblies are shown that were not considered in this report.) 
52 
53 

7 MODERATOR DATA 

This section documents the reactor moderator data. These data are important to the analysis of the 

assembly decay heat because the moderator density, soluble boron levels, and temperature affect the 

neutron spectrum in the fuel and therefore influence the average cross sections that determine the isotopic 

inventories in the fuel. The data provided in this section include the moderator soluble boron 

concentrations, density, temperature, and axial void profiles. 

7.1 Boron Levels

In PWR plants, an important moderator parameter is the soluble boron level. During a reactor operating 

cycle the boron level in the moderator is gradually reduced from its initial value at the start of a reactor 

cycle for the purposes of reactivity control. The cycle-average soluble boron levels in the Ringhals 2 and 

3 PWR plants were provided for most of the cycles. A more detailed time-dependent representation of 

the variation in boron levels can be reasonably approximated as a linearly decreasing function with a 

beginning-of-cycle value of 1.9 times the average value, and an end-of-cycle value of 0.1 times the 

average boron level. The cycle-average boron concentrations in Ringhals 2 and 3 plants are given by 

operating cycle in Table 41. 

Table 41 Average soluble boron concentrations for Ringhals 2 and 3 

Ringhals 2 Ringhals 3 

Cycle Boron (ppm) Cycle Boron (ppm) 

1 1A 

2 1B 270 

3 342 2 389 

4 530 3 452 

5 350 4 402 

6 395 5 401 

7 308 

8 323 

9 299 

10 340 

11 277 

12 302 
54 

7.2 Void Profiles 

Moderator void fractions as a function of assembly elevation are available for many, but not all, of the 

BWR assemblies. The void-fraction data for the 8 × 8, 9 × 9, SVEA-64, and SVEA-100 assemblies are 

listed in Tables 41, 42, 43 and 44, respectively. The BWR void fractions are provided at 25 equally 

spaced axial zones, consistent with the axial burnup profile data. In general, the void fractions range from 

about 0 to 0.8. In some cases negative void fraction values were reported. These negative values are 

small, and it was assumed that they are the result of predictive modeling. In the tables that follow, the 

negative void fraction values have been set to zero. 

The void profiles represent averages over the exposure history of the assembly. Power level variations 

over the course of burnup will be reflected in void-fraction variations. Following the void-fraction tables, 

there is a section discussing trends in the void profile data. 

7.3 Pressure-Density-Temperature

Moderator temperature and pressure data for each reactor are listed in Table 1. It is evident from the table 

that the operating temperatures and pressures for the BWR plants are very similar. The operating data for 

the PWR plants are also similar. The typical operating temperature and pressures were used to derive 

representative water moderator density values for each reactor type. For the PWR plants operating at a 

pressure of about 15 MPa and average temperatures of 577 K, the average water moderator density in the 

core is estimated to be about 0.72 g/cm3

. 

The BWR plants, operating at a nominal pressure of 7 MPa and a temperature of about 552 K, the density 

of the water moderator without void is about 0.75 g/cm3

. For the BWR assemblies, the effect of axial 

voiding in the reactor must also be taken into account because this reduces the effective moderator 

density. However, non-boiling regions of the assembly (e.g., water rods, water outside the box) can be 

represented using the derived value of 0.75 g/cm3

. For boiling water regions in the assembly, the 

effective density can be estimated from the void data as ρ = ρ0 (1 – V) where ρ0 = 0.75 g/cm3

 and V is the 

void fraction. Effective axial water moderator density data were provided for 11 assemblies from the 

Forsmark plants and are listed in Table 46. The effective density data estimated using the expression for 

ρ was found to agree with the actual Forsmark plant data values to within about 5%. 
Table 42 Void fractions for 8 × 8 assemblies 

Reactor Ringhals 1 Ringhals 1 Barsebäck 1 Barsebäck 1 Barsebäck 2 Oskarshamn 3 Forsmark 1 

Assembly ID 1177 1186 9329 10288 14076 12078 3838 

Axial level Void fraction 

1 (bottom) 0.000* 0.000* 0.000 0.000 0.000 0.000* 0.000* 2 0.000* 0.001 0.000 0.000 0.000 0.000* 0.000* 3 0.017 0.022 0.000 0.000 0.000 0.010 0.000* 4 0.055 0.060 0.029 0.052 0.052 0.072 0.000* 5 0.104 0.105 0.084 0.118 0.125 0.144 0.028 6 0.159 0.155 0.148 0.190 0.200 0.215 0.062 7 0.217 0.204 0.213 0.256 0.271 0.280 0.097 8 0.272 0.253 0.275 0.314 0.333 0.338 0.132 9 0.324 0.297 0.332 0.368 0.390 0.389 0.174 

10 0.371 0.337 0.384 0.419 0.441 0.436 0.221 11 0.415 0.373 0.430 0.464 0.486 0.478 0.267 12 0.453 0.406 0.472 0.504 0.525 0.514 0.309 13 0.487 0.434 0.509 0.539 0.561 0.548 0.348 14 0.516 0.461 0.543 0.571 0.592 0.579 0.386 15 0.544 0.487 0.574 0.599 0.619 0.608 0.421 16 0.569 0.511 0.601 0.634 0.642 0.633 0.452 17 0.591 0.533 0.625 0.646 0.664 0.655 0.479 18 0.611 0.554 0.647 0.666 0.683 0.675 0.504 19 0.630 0.573 0.667 0.683 0.700 0.692 0.526 20 0.646 0.589 0.683 0.699 0.715 0.707 0.544 21 0.661 0.604 0.698 0.712 0.728 0.720 0.560 22 0.674 0.617 0.712 0.734 0.739 0.732 0.574 23 0.684 0.627 0.722 0.734 0.749 0.742 0.586 24 0.693 0.635 0.730 0.741 0.756 0.749 0.594 

25 (top) 0.698 0.638 0.736 0.746 0.760 0.753 0.605 

* Reported negative void fractions set to zero. 

55 
Table 42 Void fractions for 8 × 8 assemblies (continued) 

Reactor Oskarshamn 2 

Assembly ID 1377 1389 1546 1696 1704 2995 6350 

Axial level Void fraction 

1 (bottom) 0.000 0.000 0.000 0.000 0.000 0.000 0.000* 2 0.007 0.000 0.004 0.001 0.000 0.005 0.001 3 0.040 0.002 0.027 0.011 0.006 0.028 0.038 4 0.091 0.018 0.068 0.038 0.029 0.073 0.102 5 0.146 0.045 0.114 0.078 0.068 0.129 0.171 6 0.221 0.077 0.172 0.118 0.111 0.192 0.241 7 0.298 0.113 0.232 0.162 0.162 0.253 0.304 8 0.374 0.155 0.283 0.208 0.218 0.311 0.356 9 0.435 0.198 0.334 0.250 0.272 0.363 0.402 

10 0.484 0.238 0.383 0.285 0.320 0.410 0.445 11 0.526 0.278 0.426 0.318 0.364 0.452 0.483 12 0.562 0.316 0.463 0.351 0.406 0.488 0.515 13 0.592 0.353 0.497 0.380 0.440 0.520 0.544 14 0.618 0.387 0.529 0.409 0.472 0.548 0.571 15 0.642 0.419 0.557 0.437 0.501 0.575 0.594 16 0.663 0.448 0.582 0.462 0.526 0.598 0.617 17 0.680 0.473 0.604 0.485 0.548 0.618 0.637 18 0.696 0.496 0.625 0.507 0.568 0.638 0.654 19 0.710 0.517 0.644 0.527 0.588 0.654 0.669 20 0.721 0.535 0.661 0.543 0.603 0.669 0.685 21 0.730 0.551 0.675 0.558 0.617 0.683 0.697 22 0.735 0.564 0.687 0.572 0.630 0.694 0.707 23 0.739 0.575 0.697 0.582 0.639 0.703 0.716 24 0.745 0.583 0.704 0.589 0.647 0.710 0.723 

25 (top) 0.750 0.590 0.711 0.597 0.654 0.716 0.726 

* Reported negative void fractions set to zero. 

56 
57 

Table 43 Void fractions for 9 × 9 assemblies 

Reactor Forsmark 1 

Assembly ID KU0269 KU0278 KU0282 

Axial level Void fraction 

1 (bottom) 0.000* 0.000* 0.000*

2 0.000* 0.000* 0.000*

3 0.000* 0.000* 0.000*

4 0.047 0.022 0.037 

5 0.128 0.083 0.105 

6 0.208 0.147 0.170 

7 0.283 0.208 0.236 

8 0.349 0.265 0.298 

9 0.408 0.320 0.354 

10 0.461 0.371 0.405 

11 0.507 0.418 0.450 

12 0.548 0.461 0.491 

13 0.583 0.500 0.529 

14 0.615 0.537 0.563 

15 0.643 0.569 0.594 

16 0.668 0.597 0.621 

17 0.689 0.622 0.645 

18 0.708 0.645 0.666 

19 0.725 0.664 0.684 

20 0.739 0.681 0.700 

21 0.752 0.697 0.713 

22 0.763 0.710 0.725 

23 0.771 0.720 0.733 

24 0.776 0.729 0.740 

25 (top) 0.776 0.728 0.738 

* Reported negative void fractions set to zero. 
58 

Table 44 Void fractions for SVEA-64 assemblies 

Reactor Oskarshamn 2 Forsmark 2 Forsmark 2 Forsmark 2 

Assembly ID 12684 11494 11495 13775 

Axial level Void fraction 

1 (bottom) 0.000* 0.000* 0.000* 0.000* 

2 0.000* 0.000* 0.000* 0.000* 

3 0.014 0.000* 0.000* 0.000* 

4 0.085 0.043 0.043 0.050 

5 0.167 0.114 0.114 0.122 

6 0.245 0.187 0.187 0.197 

7 0.313 0.257 0.257 0.267 

8 0.371 0.318 0.318 0.330 

9 0.423 0.374 0.374 0.385 

10 0.471 0.425 0.425 0.436 

11 0.514 0.472 0.472 0.482 

12 0.551 0.512 0.512 0.520 

13 0.584 0.548 0.548 0.555 

14 0.614 0.580 0.580 0.585 

15 0.640 0.609 0.609 0.612 

16 0.662 0.634 0.634 0.635 

17 0.683 0.657 0.657 0.656 

18 0.701 0.677 0.677 0.675 

19 0.718 0.695 0.695 0.692 

20 0.732 0.711 0.711 0.706 

21 0.744 0.725 0.725 0.719 

22 0.754 0.737 0.737 0.730 

23 0.763 0.747 0.747 0.738 

24 0.768 0.753 0.753 0.744 

25 (top) 0.773 0.759 0.759 0.747 

* Reported negative void fractions set to zero. 
59 

Table 45 Void fractions for SVEA-100 assemblies 

Reactor Oskarshamn 3 Oskarshamn 3 Forsmark 3 Forsmark 3 

Assembly ID 13628 13630 13847 13848 

Axial level Void fraction 

1 (bottom) 0.000* 0.000* 0.000* 0.000* 

2 0.000* 0.000* 0.000* 0.000* 

3 0.000* 0.001 0.000* 0.000* 

4 0.064 0.065 0.062 0.062 

5 0.145 0.146 0.143 0.143 

6 0.223 0.225 0.222 0.222 

7 0.291 0.296 0.296 0.296 

8 0.350 0.358 0.361 0.361 

9 0.401 0.414 0.418 0.418 

10 0.447 0.463 0.468 0.468 

11 0.487 0.506 0.512 0.512 

12 0.522 0.544 0.550 0.550 

13 0.554 0.578 0.583 0.583 

14 0.583 0.607 0.612 0.612 

15 0.609 0.634 0.638 0.638 

16 0.632 0.656 0.661 0.661 

17 0.653 0.677 0.682 0.682 

18 0.672 0.696 0.700 0.700 

19 0.689 0.712 0.717 0.717 

20 0.704 0.727 0.731 0.731 

21 0.717 0.739 0.744 0.744 

22 0.728 0.750 0.755 0.755 

23 0.736 0.758 0.764 0.764 

24 0.742 0.764 0.771 0.771 

25 (top) 0.742 0.763 0.771 0.771 

* Reported negative void fractions set to zero. 
Table 46 Water moderator densities for Forsmark assemblies 

Reactor Forsmark 1 Forsmark 1 Forsmark 1 Forsmark 1 Forsmark 1 Forsmark 2 Forsmark 2 Forsmark 2 Forsmark 2 

Assembly ID 3838 KU0100 KU0269 KU0278 KU0282 11494 11495 13775 5535 

Axial level Water density (g/cm3) 

1 (bottom) 0.764 0.763 0.764 0.763 0.763 0.762 0.762 0.762 0.765 2 0.762 0.760 0.760 0.760 0.760 0.756 0.756 0.757 0.762 3 0.754 0.750 0.740 0.748 0.742 0.738 0.738 0.739 0.749 4 0.738 0.722 0.700 0.720 0.706 0.703 0.703 0.707 0.724 5 0.717 0.678 0.647 0.681 0.661 0.658 0.658 0.665 0.691 6 0.693 0.625 0.594 0.639 0.617 0.611 0.611 0.621 0.659 7 0.669 0.573 0.544 0.598 0.574 0.565 0.565 0.579 0.632 8 0.643 0.525 0.498 0.560 0.535 0.523 0.523 0.540 0.613 9 0.613 0.482 0.457 0.523 0.497 0.484 0.484 0.503 0.598 

10 0.579 0.444 0.420 0.488 0.462 0.448 0.448 0.469 0.584 11 0.547 0.410 0.388 0.456 0.431 0.416 0.416 0.438 0.571 12 0.516 0.381 0.360 0.426 0.403 0.388 0.388 0.410 0.559 13 0.488 0.356 0.336 0.398 0.377 0.364 0.364 0.385 0.545 14 0.461 0.333 0.315 0.373 0.353 0.342 0.342 0.363 0.529 15 0.436 0.313 0.296 0.351 0.332 0.323 0.323 0.343 0.512 16 0.415 0.296 0.279 0.331 0.313 0.306 0.306 0.326 0.499 17 0.395 0.280 0.265 0.314 0.297 0.290 0.290 0.310 0.486 18 0.378 0.266 0.252 0.298 0.282 0.277 0.277 0.296 0.474 19 0.363 0.253 0.241 0.285 0.270 0.264 0.264 0.284 0.464 20 0.350 0.242 0.231 0.273 0.259 0.253 0.253 0.273 0.456 21 0.338 0.232 0.222 0.262 0.250 0.244 0.244 0.263 0.449 22 0.329 0.224 0.214 0.253 0.243 0.235 0.235 0.255 0.443 23 0.321 0.217 0.208 0.246 0.237 0.228 0.228 0.249 0.438 24 0.315 0.211 0.203 0.239 0.232 0.223 0.223 0.243 0.432 

25 (top) 0.308 0.208 0.204 0.239 0.233 0.219 0.219 0.239 0.418 

60 
61 

7.4 Trends in Void-Profile Data

In the absence of void profiles for some BWR assemblies, it is possible to estimate the void profile from 

the data that are available. Figures 7 and 8 show reported void profiles for selected assemblies from the 

Barsebäck 1 and Oskarshamn 2 plants, respectively. 

0.0

0.20

0.40

0.60

0.80

0 0.2 0.4 0.6 0.8 1

9329

10288

Void fraction

Fractional height

Figure 7 Void fraction profile for Barsebäck 1 assemblies 
62 

0.0

0.20

0.40

0.60

0.80

0 0.2 0.4 0.6 0.8 1

1377

1389

1546

1696

1704

2995

6350

12684

Void fraction

Fractional height

Figure 8 Void fraction profile for Oskarshamn 2 assemblies 

The shape of all the profiles is similar. In the case of the Oskarshamn 2 data, the void fractions shown 

below zero can be assumed as effectively zero and are attributed to the core simulation codes. The 

profiles for other BWR assemblies are similar. A distinguishing characteristic of the profiles is the 

maximum void fraction (or alternatively, the average void profile value). 

Void conditions are dependent on the power density in the fuel. The correlation between maximum void 

fraction and power density for the assemblies for which void data were reported is illustrated in Figure 9. 

The relationship between void and assembly power is clear. The points show some scatter as one might 

expect given that the power and void fractions used to construct the plot are averages over a number of 

burnup cycles. The best-fit trend line shown in Figure 9 is as follows: 

V = 0.522 + (0.00992)·P, 

where V is the maximum void fraction, and P is the average assembly power in MW/MTU. 

Therefore, knowing the average assembly power, one can estimate with some confidence the maximum 

void fraction experienced by the assembly. The void profiles can be obtained from assemblies from the 

same reactor having a similar power density. 
63 

0.00

0.20

0.40

0.60

0.80

1.00

0 5 10 15 20 25 30

Maximum Void Fraction

Average Power (MW/t)

Figure 9 Maximum void fraction vs. assembly average power 
64 
65 

8 SUMMARY 

This report is a compilation of assembly design data, fuel data, and reactor operating history data 

necessary for the computational analysis of decay heat measurements performed on spent fuel assemblies 

at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel, CLAB. The design data and 

measurement results are provided to enable validation of computation methods and data for the prediction 

of the heat generation in nuclear fuel. This report contains design and measurement data for 34 BWR 

assemblies and 30 PWR assemblies and includes many modern fuel designs with moderate enrichment 

levels and burnups that extend to about 51,000 MWd/MTU. The experimental data presented in this 

report expands significantly the range of validation for decay heat predictions well beyond the range 

covered by previous measurements.5,8

For some assemblies, incomplete data have been reported. Most notably, enrichment zoning and burnable 

poison rod configurations were not available for the BWR assemblies. For these data, representative 

values can generally be obtained from alternate sources. 

The measurements documented in this report include all those completed to date at the CLAB facility for 

spent fuel assemblies that were not rebuilt (removal or reconfiguration of fuel rods). The data for rebuilt 

assemblies may be added in the future as more data on the details of the reconstruction become available. 

This validation database may be expanded in the future as additional measurements are performed and 

reported for assemblies with higher enrichment and burnup values. 
66 
67 

9 REFERENCES 

1. SKB Website http://www.skb.se/default2____16807.aspx

2. SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing 

Evaluations, ORNL/TM-2005/39, Version 5, Vols. I–III, April 2005. Available from Radiation 

Safety Information Computational Center at Oak Ridge National Laboratory as CCC-725. 

3. Regulatory Guide 3.54 – Spent Fuel Heat Generation in an Independent Spent Fuel Storage 

Installation, Revision 1, U.S. Nuclear Regulatory Commission, January 1999. 

4. F. Sturek, L. Agrenius, and O. Osifo, Measurements of Decay Heat in Spent Nuclear Fuel at the 

Swedish Interim Storage Facility, Clab, Svensk Kärnbränslehantering AB, R-05-62, December 2006. 

5. B. F. Judson, In-Plant Test Measurements for Spent Fuel Storage at Morris Operation; Volume 3: 

Fuel Bundle Heat Generation Rates, General Electric report, NEDG-24922-3, February 1982. 

6. M. A. McKinnon, C. M. Heeb, and J. M. Creer, Decay Heat Measurements and Predictions of BWR 

Spent Fuel, Electric Power Research Institute report, EPRI NP-4619, June 1986. 

7. I. C. Gauld, G. Ilas, B. D. Murphy, and C. F. Weber, Validation of SCALE 5 for Decay Heat 

Predictions of LWR Spent Fuel, NUREG/CR-6972 (ORNL/TM-2008/015), prepared for the 

U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn. 

8. O. W. Hermann, C. V. Parks, and J. P. Renier, Technical Support for a Proposed Decay Heat Guide 

Using SAS2H/ORIGEN-S Data, NUREG/CR-5625 (ORNL-6698), prepared for the U.S. Nuclear 

Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., September 1994. 

9. O. W. Hermann, S. M. Bowman, M. C. Brady, and C. V. Parks, Validation of the SCALE System 

PWR Spent Fuel Isotopic Composition Analyses, ORNL/TM-12667, Lockheed Martin Energy 

Research Corporation, Oak Ridge National Laboratory, March 1995. 

10. M. D. DeHart and O. W. Hermann, An Extension of the Validation of SCALE (SAS2H) Isotopic 

Prediction for PWR Spent Fuel, ORNL/TM-13317, Lockheed Martin Energy Research Corporation, 

Oak Ridge National Laboratory, September 1996. 

11. C. E. Sanders and I. C. Gauld, Isotopic Analysis of High-Burnup PWR Spent Fuel Samples From the 

Takahama-3 Reactor, NUREG/CR-6798 (ORNL/TM-2001/259), prepared for the U.S. Nuclear 

Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., January 2003. 

12. World Nuclear Industry Handbook, Nuclear Engineering International 2005 (and previous issues). 

13. J. C. Wagner and C. V. Parks, Parametric Study of the Effect of Burnable Poison Rods for PWR 

Burnup Credit, NUREG/CR-6761 (ORNL/TM-2000/373), prepared for the U.S. Nuclear Regulatory 

Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., March 2002.
68 
A-1 

APPENDIX A: ASSEMBLY DIAGRAMS 

Diagrams (cross-section views) are shown for the various configurations of assemblies measured in the 

CLAB decay heat project in Figures A.1 through A.6. Each diagram shows the typical configuration for 

each assembly. Details of the dimensions vary from assembly to assembly within a configuration. When 

studying the diagrams, the reader can substitute the appropriate dimensions that are listed in the tables 

presented in this report. Dimensions are not always exactly to scale. This is particularly true, for 

instance, in the case of fuel cladding where the thicknesses are exaggerated for purposes of visual clarity. 

For the BWR assemblies the fuel rods are zoned according to enrichment levels, in general. This 

enrichment zoning is indicated for some of the assemblies. The lower enrichment zones (typically at the 

outer edge of the assembly), are indicated by lighter colors. The assembly enrichments listed in this 

report are assembly average values. The actual enrichment zoning information was not available at the 

time this report was published. 

For the BWR assemblies shown here, the orientation is such that the control blade will be to the top left of 

the diagram. 
A-2 

A.1 PWR Assembly Configurations

Two diagrams are shown for the PWR assemblies: a 17 × 17 (Figure A.1) assembly and a 

15 × 15 assembly (Figure A.2). The 17 × 17 assembly has 25 water holes. These correspond with 

24 guide tubes and one central instrument tube. The 15 × 15 assembly shows a typical arrangement with 

21 water holes corresponding to 20 guide tubes and 1 central instrument tube. 

Figure A.1 17 × 17 PWR assembly layout 
A-3 

Figure A.2 15 × 15 PWR assembly layout 
A-4 

A.2 BWR Assembly Configurations

The 8 × 8 assembly is shown in Figure A.3 together with the associated surrounding channel water. The 

outermost dimension is equal to the assembly pitch. Four gadolinium-bearing burnable poison rods are 

identified in yellow. The location of the poison rods was not provided, and the configuration shown was 

assumed. There is a single water rod located near the center of the assembly. The individual lattice cells 

(square) are shown. There is an additional layer of water inside of the Zircaloy box when one accounts 

for the area of the 8 × 8 lattice inside of the flow tube area. As noted, many 8 × 8 assemblies have 

slightly smaller corner rods; however, for the assembly shown here, all rods are of equal size. 

Figure A.3 8 × 8 BWR assembly layout 
A-5 

A 9 × 9 assembly, shown here with five water rods offset from the center of the assembly, is illustrated in 

Figure A.4. The fuel rod enrichment zoning is shown, with decreasing levels indicated by lighter colors. 

Enrichment zoning was not available for most of the assemblies; however the zoning pattern shown in 

Figure A.4 is typical of many BWR assemblies. 

Figure A.4 9 × 9 BWR assembly layout 
A-6 

The following two diagrams (Figures A.5 and A.6) show a SVEA-64 and a SVEA-100 assembly. The 

SVEA-64 assemblies contained one water rod; the SVEA-100 assemblies had no water rods. For both 

assemblies, typical enrichment-zoning patterns are shown with lower enrichments indicated by lighter 

colored rods toward the assembly edges. Again, detailed enrichment zoning was not available for all 

assemblies. Five gadolinium-bearing poison rods are shown in the SVEA-64 assembly, and eight are 

shown in the SVEA-100 assembly. 

Figure A.5 SVEA-64 assembly layout 
A-7 

Figure A.6 SVEA-100 assembly layout 
A-8 
B-1 

APPENDIX B: MEASURED DECAY HEAT VALUES 

The decay-heat measurement data are listed in Tables B.1 and B.2 for PWR and BWR assemblies, 

respectively. Heat generation values measured by calorimetry are listed, as well as gamma-ray escape 

heat values. Also listed are assembly discharge dates, decay heat measurement dates, cooling times and 

assembly burnup values. Note that multiple measurements were performed on some of the assemblies. 
B-2 

Table B.1 Decay heat measurement results for PWR assemblies 

Assembly Design Discharge 

datea

Measurement 

date 

Decay 

time 

(days) 

Burnup 

(MWd/MTU) 

Calorimeter 

decay heat 

(W) 

Gamma 

escape 

heat 

(W) 

Total 

decay 

heat 

(W) 

0C9 17 × 17 5/30/1986 5/6/2004 6551 38,442 478.6 12.6 491.2 

0E2 17 × 17 7/7/1988 6/16/2004 5823 41,628 572.4 15.5 587.9 

0E6 17 × 17 7/7/1988 6/22/2004 5829 35,993 475.1 12.6 487.8 

1C2 17 × 17 5/30/1986 5/14/2004 6559 33,318 407.3 10.4 417.7 

1C5 17 × 17 5/30/1986 6/17/2004 6593 38,484 486.5 12.7 499.2 

1E5 17 × 17 7/7/1988 6/11/2004 5818 34,638 456.9 11.9 468.8 

2A5 17 × 17 5/11/1984 5/3/2004 7297 20,107 227.9 5.8 233.8 

2C2 17 × 17 5/30/1986 5/5/2004 6550 36,577 454.8 11.7 466.5 

3C1 17 × 17 5/30/1986 4/30/2004 6545 36,572 458.3 11.9 470.2 

3C5 17 × 17 5/30/1986 4/28/2004 6543 38,373 488.9 12.5 501.4 

3C9 17 × 17 5/30/1986 5/7/2004 6552 36,560 456.5 11.9 468.4 

4C4 17 × 17 5/30/1986 5/27/2004 6572 33,333 411.4 10.6 422.0 

4C7 17 × 17 5/30/1986 5/4/2004 6549 38,370 485.9 12.8 498.7 

5A3 17 × 17 5/11/1984 6/13/2003 6972 19,699 231.9 5.8 237.7 

5A3 17 × 17 5/11/1984 6/16/2003 6975 19,699 230.8 5.8 236.7 

5A3 17 × 17 5/11/1984 6/18/2003 6977 19,699 237.6 5.8 243.4 

5A3 17 × 17 5/11/1984 4/27/2004 7291 19,699 225.1 5.8 230.9 

5A3 17 × 17 5/11/1984 5/10/2004 7304 19,699 224.4 5.8 230.3 

a

 Dates listed as month/day/year. 
B-3 

Table B.1 Decay heat measurement results for PWR assemblies (continued)

Assembly Design Discharge 

datea

Measurement 

date 

Decay 

time 

(days) 

Burnup 

(MWd/MTU) 

Calorimeter 

decay heat 

(W) 

Gamma 

escape 

heat 

(W) 

Total 

decay 

heat 

(W) 

C01 15 × 15 4/4/1981 6/10/2004 8468 36,688 406.3 9.4 415.8 

C12 15 × 15 4/4/1981 4/6/2004 8403 36,385 401.6 8.7 410.3 

C20 15 × 15 4/4/1985 4/14/2004 6950 35,720 406.1 9.7 415.8 

C20 15 × 15 4/4/1985 4/15/2004 6951 35,720 416.4 9.6 426.0 

C20 15 × 15 4/4/1985 4/16/2004 6952 35,720 419.2 9.7 428.9 

C20 15 × 15 4/4/1985 4/23/2004 6959 35,720 425.9 9.8 435.7 

D27 15 × 15 4/28/1983 4/26/2004 7669 39,676 444.6 11.4 456.1 

D38 15 × 15 5/6/1982 4/5/2004 8005 39,403 431.6 10.8 442.3 

E38 15 × 15 5/6/1982 3/30/2004 7999 33,973 366.9 9.4 376.3 

E38 15 × 15 5/6/1982 3/31/2004 8000 33,973 364.9 9.4 374.3 

E40 15 × 15 5/6/1982 6/14/2004 8075 34,339 372.2 9.0 381.3 

F14 15 × 15 4/28/1983 6/18/2004 7722 34,009 372.5 9.3 381.8 

F21 15 × 15 4/13/1984 6/23/2004 7376 36,273 410.2 10.7 420.9 

F25 15 × 15 4/28/1983 6/21/2004 7725 35,352 386.7 10.0 396.7 

F32 15 × 15 5/12/1988 5/28/2004 5860 50,962 675.4 16.6 692.0 

G11 15 × 15 4/4/1985 5/24/2004 6990 35,463 405.7 10.6 416.4 

G23 15 × 15 4/4/1985 5/18/2004 6984 35,633 409.6 11.1 420.6 

I09 15 × 15 5/12/1988 5/17/2004 5849 40,188 494.5 13.4 507.9 

I24 15 × 15 4/30/1986 5/26/2004 6601 34,294 398.2 11.8 410.1 

I25 15 × 15 4/25/1987 4/13/2004 6198 36,859 434.0 11.8 445.8 a

 Dates listed as month/day/year.
B-4 

Table B.2 Decay heat measurement results for BWR assemblies 

Assembly Design Discharge 

date 

Measurement 

date 

Decay 

time 

(days) 

Burnup 

(MWd/MTU) 

Calorimeter 

 decay heat 

(W) 

Gamma 

escape 

heat 

(W) 

Total

 decay

 heat

 (W) 

1177 8 × 8 8/2/1985 11/26/2003 6690 36,242 173.4 4.5 177.9 

1186 8 × 8 8/2/1985 11/10/2003 6674 30,498 137.5 3.3 140.8 

1377 8 × 8 5/13/1977 1/22/2004 9750 14,546 54.8 1.4 56.2 

1389 8 × 8 7/15/1981 11/28/2003 8171 19,481 81.7 2.2 83.9 

1546 8 × 8 8/19/1983 1/16/2004 7455 24,470 105.4 2.7 108.1 

1696 8 × 8 8/19/1983 12/3/2003 7411 20,870 90.0 2.3 92.4 

1704 8 × 8 7/23/1982 12/8/2003 7808 19,437 81.8 2.1 84.0 

2995 8 × 8 7/15/1981 1/7/2004 8211 29,978 127.3 3.2 130.5 

3838 8 × 8 7/10/1992 12/10/2003 4170 25,669 123.4 3.4 126.8 

3838 8 × 8 7/10/1992 12/11/2003 4171 25,669 122.6 3.4 125.9 

5535 8 × 8 7/15/1988 12/18/2003 5634 19,944 82.4 2.2 84.6 

6350 8 × 8 6/7/1985 12/4/2003 6754 27,675 123.7 3.2 126.9 

6350 8 × 8 6/7/1985 12/5/2003 6755 27,675 126.2 3.2 129.4 

6423 8 × 8 8/5/1988 2/12/2004 5669 35,109 169.1 5.2 174.2 

6432 8 × 8 8/5/1988 6/10/2003 5422 36,861 179.9 5.5 185.5 

6432 8 × 8 8/5/1988 6/12/2003 5424 36,861 184.0 5.5 189.5 

6432 8 × 8 8/5/1988 2/13/2004 5670 36,861 178.9 5.5 184.4 

6432 8 × 8 8/5/1988 2/23/2004 5680 36,861 177.1 5.7 182.8 

6432 8 × 8 8/5/1988 2/24/2004 5681 36,861 179.4 5.7 185.2 

6432 8 × 8 8/5/1988 3/1/2004 5687 36,861 175.8 5.8 181.6 

6432 8 × 8 8/5/1988 3/2/2004 5688 36,861 176.4 5.5 182.0 

6454 8 × 8 8/15/1986 2/17/2004 6395 37,236 181.3 5.0 186.3 

8327 8 × 8 8/6/1991 3/10/2004 4600 37,851 191.0 6.0 196.9 

8331 8 × 8 9/15/1989 3/11/2004 5291 35,903 181.3 5.7 187.0 

8332 8 × 8 8/5/1988 3/4/2004 5690 34,977 163.0 5.1 168.1 

8338 8 × 8 8/5/1988 3/9/2004 5695 34,830 164.9 4.6 169.5 
B-5 

Table B.2 Decay heat measurement results for BWR assemblies (continued)

Assembly Design Discharge 

date 

Measurement 

date 

Decay 

time 

(days) 

Burnup 

(MWd/MTU) 

Calorimeter 

decay heat 

(W) 

Gamma 

escape 

heat 

(W) 

Total 

decay 

heat 

(W) 

9329 8 × 8 9/17/1988 6/2/2003 5371 41,127 217.4 5.4 222.8

9329 8 × 8 9/17/1988 6/4/2003 5373 41,127 218.9 5.4 224.4

9329 8 × 8 9/17/1988 11/20/2003 5542 41,127 213.2 5.4 218.7

10288 8 × 8 9/17/1988 11/12/2003 5534 35,218 181.2 4.6 185.8

11494 SVEA-64 7/15/1988 12/2/2003 5618 32,431 161.7 4.4 166.0

11495 SVEA-64 7/15/1988 11/7/2003 5593 32,431 163.3 4.3 167.6

12078 8 × 8 7/8/1988 11/18/2003 5611 25,160 116.9 3.3 120.2

12684 SVEA 64 8/2/1991 12/16/2003 4519 46,648 274.9 7.8 282.7

13628 SVEA-100 6/24/1991 1/8/2004 4581 35,619 188.2 5.8 194.0

13630 SVEA-100 6/24/1991 12/12/2003 4554 40,363 228.6 7.1 235.7

13775 SVEA-64 7/12/1991 12/19/2003 4543 32,837 173.1 5.3 178.4

13847 SVEA-100 7/13/1990 11/13/2003 4871 31,275 165.4 4.9 170.3

13847 SVEA-100 7/14/1990 11/14/2003 4871 31,275 164.7 4.9 169.6

13848 SVEA-100 7/13/1990 11/24/2003 4882 31,275 165.6 5.1 170.7

14076 8 × 8 7/2/1992 12/9/2003 4177 40,010 233.7 6.6 240.3

KU0100 8 × 8 8/17/1990 1/9/2004 4893 34,193 180.2 5.1 185.3

KU0269 9 × 9 8/17/1990 1/19/2004 4903 35,113 187.1 5.6 192.7

KU0278 9 × 9 5/24/1991 12/22/2003 4595 35,323 189.9 5.5 195.4

KU0282 9 × 9 5/24/1991 12/1/2003 4574 37,896 212.5 5.9 218.5
NUREG/CR-6971 

ORNL/TM-2008/16 

 December 2009