Spent nuclear fuel for disposal in the KBS-3 repository

Technical Report

TR-10-13

Spent nuclear fuel for disposal
in the KBS-3 repository
Svensk Kärnbränslehantering AB

December 2010

Svensk Kärnbränslehantering AB

Box 250, SE-101 24 Stockholm
Phone +46 8 459 84 00

TR-10-13

CM Gruppen AB, Bromma, 2010

Swedish Nuclear Fueland Waste Management Co

ISSN 1404-0344
Tänd ett lager:
SKB TR-10-13
P, R eller TR.

Spent nuclear fuel for disposal
in the KBS-3 repository
Svensk Kärnbränslehantering AB

December 2010

Keywords: SKBdoc id 1175231, Spent nuclear fuel, Safety report, Fuel quantities,
Handling, Encapsulation, Decay power, Criticality, Initial state, Radionuclide
inventory.
A pdf version of this document can be downloaded from www.skb.se.

Preface
An important part of SKB’s licence application for the construction, possession and operation of
the KBS-3 repository is the safety report. The safety report addresses both safety during operationof the KBS-3 repository facility (SR‑Operation), and the long-term safety of the KBS-3 repository(SR‑Site).
For the construction of the KBS‑3 repository SKB has defined a set of production lines:•

the spent nuclear fuel,•the canister,•

the buffer,•

the backfill,•

the closure, and•the underground openings.

These production lines are reported in separate Production reports, and in addition there is a
Repository production report presenting the common basis for the reports.
This set of reports addresses design premises, reference design, conformity of the reference design
to design premises, production and the initial state, i.e. the results of the production. Thus the reports
provide input to SR‑Site concerning the characteristics of the as built KBS‑3 repository and to
SR‑Operation concerning the handling of the engineered barriers and construction of undergroundopenings.
The preparation of the set of reports has been lead and coordinated by Lena Morén with support
from Roland Johansson, Karin Pers and Marie Wiborgh.
This report has been authored by Per Grahn, Lena Morén and Marie Wiborgh.

TR-10-133

Summary
The report is included in a set of Production reports, presenting how the KBS-3 repository isdesigned, produced and inspected. The set of reports is included in the safety report for the KBS-3
repository and repository facility. The report provides input to the assessment of the long-term safety,
SR-Site as well as to the operational safety report, SR-Operation.
The report presents the spent fuel to be deposited, and the requirements on the handling and selectionof fuel assemblies for encapsulation that follows from that it shall be deposited in the KBS-3 repository. An overview of the handling and a simulation of the encapsulation and the resulting canisters
to be deposited are presented. Finally, the initial state of the encapsulated spent nuclear fuel is given.
The initial state comprises the radionuclide inventory and other data required for the assessment of
the long-term safety.
Spent nuclear fuel to be deposited in the KBS-3 repository

The major part of the spent nuclear fuel to be deposited in the KBS-3 repository consists of fuelfrom the operation of the twelve Swedish nuclear power plants. The presented fuel quantities are
based on the spent fuel stored in the interim storage facility, Clab, and on SKB’s reference scenario
for the operation of the power plants. The number of BWR and PWR assemblies to be deposited,their burnup and their ages the last year of operation of the last reactor to close down in the reference
scenario, are presented.
There are also minor quantities of miscellaneous fuels, from research and the early part of theSwedish nuclear power programme to be deposited in the KBS-3 repository. The sources andamounts of these fuels are presented.
The properties of the spent nuclear fuel impact the design of the KBS-3 repository. The design of theKBS-3 repository has in turn resulted in requirements on the handling of the spent fuel. The spent
nuclear fuel dimensions, enrichment, burnup and age are properties imposing requirements on thehandling.
Requirements on the handling of the spent fuelThe decay power of a fuel assembly depends on its burnup, age and the mass of uranium. The decay
power will influence the temperature in the final repository. Since the temperature in the buffer need
to be restricted there is a maximum allowed total decay power of the assemblies in a canister.
Criticality must always be prevented in the handling of the spent fuel. The assemblies shall beselected for encapsulation with respect to their enrichment and burnup, and the design of the canister
so that criticality under no circumstances can occur in the canister.
The dimensions of the largest fuel assemblies constitute design premises for the canister. The
canister in turn imposes requirements on the handling of the fuel. Small assemblies shall if necessary
be provided with distance devices that prevent them from moving in the channel tubes of the insert.
To avoid corrosion inside the canister insert the fuel assemblies shall be dried prior to encapsulationand the air in the insert exchanged for inert gas.
With respect to the environment in the final repository there is a maximum acceptable radiation dose
rate at the canister surface. Since the radiation dose rate in similarity to the decay power depends
on the burnup and age of the encapsulated assemblies the maximum allowed decay power will also
restrict the radiation. Both with respect to long-term safety and the radiation protection during operation it shall be verified that the radiation dose rate at the canister surface does not exceed acceptablelevels.
With respect Sweden’s commitments regarding non-proliferation and the control of fissile material
each sealed canister shall be marked and constitute a unit for the account of nuclear material.

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The handling of the spent fuel and the canisters to be deposited

The spent nuclear fuel is accepted for transportation from the nuclear power plants and is delivered
to the interim storage facility. After a period of interim storage the fuel assemblies are selected for
encapsulation with respect to their burnup and age so that the decay power in the canister does notexceed the acceptable level. It is verified that criticality cannot occur in the canister for the selectedassemblies and that the radiation dose rate at the canister surface does not exceed the acceptablelevel. The spent fuel assemblies are then transferred to the encapsulation building, dried and placedin the canister. After inspection of the assembly identities the steel lid is placed on the insert and the
air inside it is changed with argon. Finally, the copper lid is placed on the canister and it is sealed.
The encapsulation of the spent fuel has been simulated based on the spent nuclear fuel to bedeposited, the planned operation times of SKB’s facilities and the requirements on selection of fuelassemblies for encapsulation. The simulation results in the number of canisters to be deposited andthe burnup of the assemblies in the canisters.
Initial state – encapsulated spent nuclear fuelEncapsulated spent nuclear fuel comprises the spent fuel assemblies and the gases and liquids in thecavities of the canister. The initial state refers to the properties of the encapsulated fuel when thecanister is finally sealed and no more handling of individual assemblies is possible.
The radionuclide inventory at the initial state is an important input to the safety assessment. The total
radionuclide inventory in the final repository ultimately depends on the total energy output from thenuclear power plants. The radionuclide inventory in individual spent fuel assemblies will mainly
depend on the burnup and consequently, since the decay power also depends on the burnup, theradionuclide inventory in a canister is restricted by the maximum allowed decay power.
The radionuclide inventory in the final repository has been calculated as the sum of the calculated
inventories in individual assemblies, and as the sum of the inventories inventories in a set of typecanisters. The type canisters are selected based on the results from the simulation of the encapsulation to provide a representative and adequate description of the canisters’ content of spent nuclear
fuel. The radionuclide inventories in the type canisters are also an input to the assessment of theoperational safety.
In addition to the radionuclide inventory the propensity for criticality decay power, encapsulatedgases and liquids, the radiation at the canister surface and other parameters of importance for theassessment of the long-term safety are reported for the initial state.

6TR-10-13

Sammanfattning
Rapporten ingår i en grupp av Produktionsrapporter som redovisar hur KBS-3-förvaret är utformat,
producerat och kontrollerat. Gruppen av rapporter ingår i säkerhetsredovisningen för KBS-3-förvaret
och förvarsanläggningen. Rapporten levererar indata till analysen av den långsiktiga säkerheten,
SR-Site, samt till redovisningen av driftsäkerheten, SR-Drift.
Rapporten redovisar det använda kärnbränsle som ska deponeras, och kraven på hantering och val
av bränsleelement för inkapsling som följer av att det ska deponeras i KBS-3-förvaret. En översikt
av hanteringen och en simulering av inkapslingen och det resulterande antalet kapslar som ska
deponeras redovisas. Slutligen redovisas initialtillståndet för det inkapslade använda kärnbränslet.
Initialtillståndet omfattar radionuklidinventariet och andra data som behövs för analysen av denlångsiktiga säkerheten.
Använt kärnbränsle som ska deponeras i KBS-3-förvaret

Det använda kärnbränsle som ska deponeras i KBS-3-förvaret utgörs till största delen av bränsle
från driften av de tolv svenska kärnkraftverken. De redovisade bränslemängderna är baserade på det
använda kärnbränsle som lagras i mellanlagtret, Clab, och på SKB:s referensscenario för driften av
kärnkraftverken. Antalet BWR och PWR-element som ska deponeras, deras utbränning och derasålder det sista driftåret för den reaktor som stängs sist i referensscenariot redovisas.
Det finns också mindre mängder av udda bränslen, från forskning och den tidiga delen av det
svenska kärnkraftsprogrammet, som ska deponeras i KBS-3-förvaret. Källor till och mängder av
dessa bränslen redovisas.
Egenskaperna hos det använda kärnbränslet påverkar utformningen av KBS-3-förvaret. Utform­
ningen av KBS-3-förvaret har i sin tur resulterat i krav på hanteringen av det använda bränslet.
Det använda kärnbränslets mått, anrikning, utbränning och ålder medför krav på hanteringen.
Krav på hanteringen av det använda kärnbränslet

Resteffekten hos ett bränsleelement beror på dess utbränning, ålder och vikt uran. Resteffektenpåverkar temperaturen i slutförvaret. Eftersom temperaturen i bufferten behöver begränsas finns
det en maximalt tillåten total resteffekt hos elementen i en kapsel.
Kriticitet måste alltid förhindras vid hanteringen av det använda bränslet. Bränsleelementen ska
väljas för inkapsling med hänsyn till sin anrikning och utbränning, och med hänsyn till kapselns
utformning, så att kriticitet under inga omständigheter kan uppstå i kapseln.
Bränsleelementens mått utgör konstruktionsförutsättningar för kapseln. Kapseln ställer i sin tur krav
på hanteringen av bränslet. Små element ska om nödvändigt förses med distansklossar som hindrar
dem från att röra sig i insatsens bränslekanaler. För att undvika korrosion inuti kapselns insats ska
bränsleelementen torkas innan de kapslas in och luften i insatsen ska bytas ut mot inert gas.
Med hänsyn till miljön i slutförvaret finns en högsta tillåten strålningsdosrat på kapselytan. Eftersomstrålningsdosraten i likhet med resteffekten beror av det inkapslade elementens utbränning och ålder
kommer den maximalt tillåtna resteffekten också att begränsa strålningen. Både med hänsyn tilllångsiktig säkerhet och strålskydd under drift ska det verifieras att strålningsdosraten på kapselytan
inte överskrider tillåtna nivåer.
Med hänsyn till Sveriges åtaganden då det gäller icke-spridning och kontroll av klyvbart material ska
varje försluten kapsel vara märkt och utgöra en enhet i redovisningen av kärnämne.

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Hanteringen av det använda kärnbränslet och de kapslar som ska deponeras

Det använda kärnbränslet godkänns för transport från kärnkraftverken och levereras till mellan­
lagret. Efter en tids mellanlagring av bränsleelementen väljs de ut för inkapsling med hänsyn
till sin utbränning och ålder så att resteffekten i kapseln inte överskrider den tillåtna nivån. Det
verifieras att kriticitet inte kan uppstå i kapseln för de valda elementen och att strålningsdosraten på
kapselytan inte överskrider den tillåtna nivån. De använda kärnbränsleelementen förs sedan över tillinkapslingsbyggnaden, torkas och placeras i kapseln. Efter kontroll av elementens identitet placerasstållocket på insatsen och luften i instasen byts mot argon. Slutligen placeras kopparlocket påkapseln och den försluts.
Inkapslingen av det använda kärnbränslet har simulerats utifrån det använda kärnbränsle som ska
deponeras, den planerade driften av SKB:s anläggningar och kraven på val av bränsleelement för
inkapsling. Simuleringen resulterar i det antal kapslar som ska deponeras och utbränningen hoselementen i kapslarna.
Initialtillstånd – inkapslat använt kärnbränsle

Inkapslat använt kärnbränsle omfattar de använda bränsleelementen och gaserna och vätskorna ikapselns hålrum. Initialtillståndet avser det inkapslade bränslets egenskaper då kapseln slutligen
försluts och ingen mer hantering av enskilda element är möjlig.
Radionuklidinventariet vid initialtillståndet utgör viktig indata till säkerhetsanalysen. Det totala
radionuklidinventariet i slutförvaret beror ytterst av den totala energiproduktionen i kärnkraftverken.
Radionuklidinventariet i enskilda element beror huvudsakligen på utbränningen och följaktligen,eftersom resteffekten också beror av utbränningen, begränsas radionuklidinnehållet i en kapsel av
den maximalt tillåtna resteffekten.
Radionuklidinventariet i slutförvaret har beräknats som summan av de beräknade inventarierna ienskilda element, och som summan av inventarierna i en uppsättning typkapslar. Typkapslarna har
valts baserat på resultaten från simuleringen av inkapslingen, för att ge en representativ och adekvat
beskrivning av kapslarnas innehåll av använt kärnbränsle. Radionuklidinventariet i typkapslarna
utgör också indata till analysen av driftsäkerheten.
Utöver radionuklidinventariet redovisas benägenheten för kriticitet, resteffekt, inkapslade gaser
och vätskor, strålningen på kapselns yta och andra parametrar med betydelse för analysen av denlångsiktiga säkerheten för initialtillståndet.

8

TR-10-13

Contents
1
1.1

1.2

1.3

1.4
2
2.1
2.2

2.3

3
3.1

3.2

4
4.1
4.2
4.3
4.4
4.5

Introduction
General basis
1.1.1 This report
1.1.2 The spent nuclear fuel to be deposited
1.1.3 The handling of the spent nuclear fuelPurpose, objectives and delimitations
1.2.1 Purpose
1.2.2 Objectives
1.2.3 Limitations
Interfaces to other reports included in the safety report
1.3.1 The safety report for the long-term safety
1.3.2 The safety report for the operational safety
1.3.3 The other production reports
Structure and content

11
11
11
11
12
12
12
12
12
12
13
13
13
14

Spent nuclear fuel to be deposited in the KBS-3 repository
Types and sources of spent nuclear fuel2.1.1 Reference scenario for the operation of the nuclear power plants
2.1.2 Miscellaneous fuels – background and sources
Fuel quantities, burnup and age of the spent nuclear fuel2.2.1 Spent fuel from the reference scenario for operation of the nuclear
power plants
2.2.2 Miscellaneous fuels
Fuel parameters of importance for repository design and long-term safety
2.3.1 Enrichment, burnup, irradiation and power history and age of the
spent fuel assemblies
2.3.2 Dimensions and materials
2.3.3 Encapsulated liquids and gases

15
15
15
16
16

Requirements on the handling of the spent fuelRequirements related to repository design and long‑term safety
3.1.1 Decay power
3.1.2 Criticality
3.1.3 Dimensions and spacing devices
3.1.4 Encapsulated liquids and gases
3.1.5 Radiation
Requirements related to the operation of the KBS-3 system
3.2.1 Encapsulation
3.2.2 Operational safety and radiation protection
3.2.3 Control of nuclear material – Safeguards

25
25
25
25
26
27
27
28
28
28
28

The handling of the spent nuclear fuelOverview
Transport and delivery of fuel assemblies
4.2.1 Activities
4.2.2 Inspections
Interim storage
4.3.1 Activities
4.3.2 Inspections
Selection of assemblies
4.4.1 Activities
4.4.2 Inspections
Delivery to the encapsulation building
4.5.1 Activities
4.5.2 Inspections

29
29
29
29
30
30
30
30
30
30
34
34
34
34

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16
19
20
21
21
23

9

4.6

Drying of fuel assemblies
4.6.1 Activities
4.6.2 Inspections
Placement in the canister
4.7.1 Activities
4.7.2 Inspections
Replacement of atmosphere in the canister insert
4.8.1 Activities
4.8.2 Inspections

35
35
35
35
35
35
36
36
36

5
5.1
5.2

The canisters to be deposited
Factors that will affect the number of canisters
Simulation of the encapsulation

37
37
38

6
6.1
6.2

41
41
41
41
42
43
45
50
51

6.4
6.5
6.6
6.7
6.8

Initial state – encapsulated spent nuclear fuelIntroduction
Radionuclide inventory
6.2.1 Fuel parameters of importance for the radionuclide inventory
6.2.2 Total radionuclide inventory in the final repository
6.2.3 The type canister approach
6.2.4 The type canisters and their radionuclide inventory
6.2.5 Comparison of inventory in type canisters with total inventory
6.2.6 Uncertainties in the calculated fuel matrix radionuclide inventories
6.2.7 Uncertainties in the calculated radionuclide inventories inconstruction materials and crud
Fission gas release and gap inventory
6.3.1 Fission gas release in the spent fuel assemblies
6.3.2 The fission gas release in the type canisters
Decay power
Encapsulated gases and liquids
Radiation at the canister surface
Criticality
Dimensions and other parameters of interest for the safety assessment

7

References

59

4.7
4.8

6.3

Appendix A Fuel types
Appendix B Material specification for representative fuel types for BWR
and PWR
Appendix C Radionuclide inventory
Appendix D Glossary of abbreviations and branch terms

10

52
53
53
56
56
56
57
57
57
61
65
71
97

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1

Introduction

1.1

General basis

1.1.1 This report
This report describes the spent nuclear fuel to be deposited in the KBS-3 repository and the handlingof the spent fuel within the KBS-3 system. It is included in a set of reports presenting how theKBS-3 repository is designed, produced and inspected. The set of reports is denominated Production
reports. The Production reports and their short names used as references within the set are illustrated
in Figure 1‑1. The reports within the set referred to in this report and their full names are presented
in Table 1‑1.
This report is part of the safety report for the KBS-3 repository and repository facility, see
Repository production report, Section 1.2. It is based on the results and review of the most recent
long-term safety assessment and the current knowledge, technology and results from research anddevelopment.

1.1.2 The spent nuclear fuel to be deposited
All spent nuclear fuel from the currently approved Swedish nuclear power programme shall bedeposited in the KBS‑3 repository. The presented amounts and types of fuel to be deposited are
based on the spent fuel already accumulated and in interim storage and a prognosis of the amounts
and types of spent fuel to be generated in a scenario for the future operation of the nuclear power
plants.

Production reports
Repository
production
reportSpent fuelreport”Engineered barrier ” production reports

Canisterproduction
reportBufferproduction
reportBackfill
production
reportClosureproduction
reportUndergroundopenings
construction
reportFigure 1‑1. The reports included in the set of reports describing how the KBS-3 repository is designed,
produced and inspected.
Table 1‑1. The reports within the set of Production reports referred to in this report.
Full title

Short name used within the
Production reports

Text in reference lists

Design and production of theKBS-3 repository

Repository production report

Repository production report, SKB 2010.
Design and production of the KBS-3 repository.
SKB TR-10-12, Svensk Kärnbränslehantering AB.

Design, production and initialstate of the canister

Canister production report

Canister production report, SKB 2010.
Design, production and initial state of the canister.
SKB TR-10-14, Svensk Kärnbränslehantering AB.

Design, construction andinitial state of the undergroundopenings

Underground openings
construction report

Underground openings construction report,
SKB 2010. Design, construction and initial stateof the underground openings. SKB TR-10-18,
Svensk Kärnbränslehantering AB.

TR-10-1311

1.1.3 The handling of the spent nuclear fuelThe presented handling of the spent fuel is based on the assumption that there is a system, the KBS-3
system, comprising the facilities required to manage the spent nuclear fuel and finally deposit it in aKBS-3 repository. The KBS-3 system and its facilities are presented in Chapter 4 in the Repository
production report.
SKB considers that the methods to handle the spent fuel and inspect its properties presented inthis report are technically feasible. Methods for handling and inspection may, however, be further
developed before the encapsulation and disposal of the fuel commences.

1.2

Purpose, objectives and delimitations

1.2.1 Purpose
The purpose of this report is to present the spent nuclear fuel to be disposed in the KBS-3 repository.
The report shall provide the information on the encapsulated spent fuel required for the long-term
safety report, SR-Site, as well as for the operational safety report, SR-Operation.
With this report SKB intends to present the requirements on the handling of the spent fuel for its
disposal in the KBS-3 repository and how it can be handled in conformity to these requirements.
The report shall present the handling and inspection and summarise the efforts that supports that the
spent fuel is handled in conformity to the stated requirements.

1.2.2 Objectives
Based on the above purpose the objectives of this report are to describe:•

spent fuel types and quantities to be deposited in the KBS‑3 repository,•

fuel properties and parameters of importance for the assessment of the long-term safety,•

how the fuel assemblies are handled, inspected and selected for encapsulation and deposition,•

the initial state of the spent fuel, i.e. the expected values of parameters of importance for theassessment of the long-term safety of the encapsulated spent nuclear fuel.

1.2.3 Limitations
The Spent fuel report is based on a reference scenario for the future operation of the nuclear
power plants and also includes the spent fuel that is stored in the interim storage facility. Alternative
scenarios for the operation of the nuclear power plants are not included.
The Spent fuel report includes data about the spent fuel that are required to assess the safety of theKBS-3 repository and repository facility. Other fuel data of interest for SKB will be documented
elsewhere.
The Spent fuel report only includes the inspections performed within the KBS-3 system thatare required to inspect parameters of importance for the assessment of the safety of the KBS-3
repository and repository facility. Inspections that are performed at the nuclear power plants, or withrespect to safeguards of nuclear materials, or operational safety of the other facilities or transportsystem within the KBS-3 system are reported elsewhere.

1.3

Interfaces to other reports included in the safety report

The role of the production reports in the safety report is presented in Section 1.2 in the Repository
production report. A summary of the interfaces to other reports included in the safety report isgiven below.

12TR-10-13

1.3.1 The safety report for the long-term safety
By providing a basic understanding of the repository performance for different time periods and by
the identification of scenarios that can be shown to be especially important from the standpoint ofrisk, the long-term safety assessment provides feedback to the allocation of spent fuel assemblies
to canisters, and the design of the engineered barriers and underground openings. The methodology
used for deriving design premises from the long-term safety assessment is introduced in the
Repository production report, Section 2.5.2. A more thorough description as well as the resulting
design premises are given in the report “Design premises for a KBS-3V repository based on resultsfrom the safety assessment SR‑Can and some subsequent analyses” /­SKB 2009/, hereinafter referred
to as Design premises long-term safety. These design premises constitute an input to the required
handling of the spent nuclear fuel.
As stated in Section 1.2 this report provides information about the initial state of the encapsulated
spent nuclear fuel required for the long-term safety assessment. This report also provides the data
concerning the initial state used in the calculations included in the long-term safety assessment.

1.3.2 The safety report for the operational safety
The spent fuel line includes the selection of fuel assemblies to be encapsulated and ends in theencapsulation plant when the fuel is put in the canister. The general description of the facility and its
main activities given in Chapter 5 in the Safety Report for the central interim storage and encapsulation plant, Clink, constitute input to this report.
The objectives for the operational safety and radiation protection in the final repository facility given
in Chapter 3 in the safety report SR-Operation constitute input to this report.
The radioactivity of the encapsulated spent fuel is also of importance for the control of radioactive
substances and radiation protection in the repository facility, and constitutes input to Chapters 6 and 7
in SR-Operation.

1.3.3 The other production reports
The Repository production report presents the context of the set of Production reports and theirrole within the safety report. It also includes definitions of some central concepts of importance forthe understanding of the Production reports.
The Repository production report provides an overview of the KBS-3 system and the facilities
required to manage and deposit the spent nuclear fuel. It also points out the properties of the spent
nuclear fuel of importance for the design of the KBS-3 repository and how the design of the KBS‑3
repository will impact the handling of the spent fuel.
The design premises the spent fuel imposes on the design and construction of the engineered barriers
and underground openings are presented in this report. These design premises are repeated andverified in the production reports for the engineered barriers and underground openings on which
the spent nuclear fuel imposes design premises.
The handling of the spent nuclear fuel and the canister production line intersect in the Clink facility.
The selection and inspection of spent fuel assemblies for each canister, and the activities taking place
before the canister is sealed are described in this report. The sealing of the canister and all activities
taking place after the sealing are presented in the Canister production report.

TR-10-1313

1.4

Structure and content

The Spent fuel report contains following chapters.
•

Spent nuclear fuel to be deposited in the KBS-3 repository
Including information about fuel types and quantities and the fuel parameters of importance forthe assessment of the long-term safety.

•

Requirements on the handling of spent nuclear fuelIncluding the requirements from the KBS-3 repository and its engineered barriers on the handlingand selection of fuel assemblies for encapsulation and the design premises imposed on theengineered barriers by the spent fuel.

•The handling of the spent nuclear fuelIncluding the activities included in the spent fuel line and the inspections and verifications of thefuel parameters of importance for the safety report.

•

The canisters to be deposited
The number of canisters to be deposited as a result of the planned handling and encapsulation.

•

Initial state – encapsulated spent nuclear fuelIncluding the expected fuel parameters of importance for the assessment of the long-term safety.

In addition to this an introduction to the report is given in this chapter and in Appendix D abbreviations and branch terms used in this report are explained.

14

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2

Spent nuclear fuel to be deposited in theKBS‑3 repository

2.1

Types and sources of spent nuclear fuelThe major part of the nuclear fuel to be deposited in the KBS‑3 repository consists of spent fuel fromthe operation of the twelve Swedish nuclear power plants. The quantities and properties of the spent
fuel from the twelve Swedish nuclear power plants are based on the scenario for their operation
presented in Section 2.1.1.
There are also minor quantities of other fuel types from research and the early part of the nuclear
power programme to be deposited in the KBS-3 repository. These fuels are in the following referred
to as miscellaneous fuels. A background to the sources of these fuels is presented in Section 2.1.2.

2.1.1 Reference scenario for the operation of the nuclear power plants
In this section the reference scenario used by SKB for the operation of the nuclear power plants ispresented. This scenario is the basis for the reference design of the final repository and the analyses
within SR-Site.
The Swedish nuclear power plants are of the type boiling light water reactors (BWR) and pressurised
light water reactors (PWR). In total there are 12 reactors; two BWR at Barsebäck (B1 and B2), three
BWR at Forsmark (F1, F2 and F3), one BWR (R1) and three PWR at Ringhals (R2, R3 and R4) andthree BWR at Oskarshamn (O1, O2 and O3).
The maximum enrichment of the spent nuclear fuel to be deposited in the KBS-3 repository is set to
5% and the average assembly burnup is limited to 60 MWd/kg U for the uranium oxide fuel (UOX
fuel) from the BWR and PWR plants and 50 MWd/kg HM1 for the mixed oxide fuel (MOX) BWR
fuel /­SKB 2008/.
The B1 and B2 reactors in Barsebäck are closed down and had been in operation for approximately
24 years and 28 years, respectively, when they were taken out of operation. The operating times are
set to 50 years for the reactors at Ringhals and Forsmark and 60 years for the reactors at Oskars­hamn.
In the reference scenario the last reactor will be taken out of operation in 2045. In the scenario for
the future operation of the power plants the thermal powers of the reactors given in Table 2‑1 areassumed. The table also includes the reactor powers of the closed-down reactors in Barsebäck.
Table 2‑1. Thermal powers of the reactors and last year of operation for the 12 Swedish nuclear
power plants as assumed in the reference scenario used by SKB. Based on information provided
by the nuclear power plant operators.
Reactor

Reactor power(MWth)

B1

1,800

B2

1,800

F1

2,928

F2

2,928

F3

3,300

O1

1,375

O2

1,800

O3

3,300

R1

2,540

R2

2,652

R3

2,992

R4

2,775

1

Increases in reactor power (MWth)
2009
2010
2011
2012

Last year ofoperation

1999
2005
3,255

2030

3,255

2031
3,775

2035
2032
2,300

3,900

2034
2045
2025
2025

3,144

2031
3,300

2033

Heavy metal i.e. uranium and plutonium.

TR-10-1315

In order to estimate the number of spent fuel assemblies generated as a result of the future operationof the nuclear power plants the planned target burnup of the fuel assemblies as well as the thermal
power must be known. The batch average discharge burnup for the ten remaining Swedish nuclear
power plants that are assumed in the reference scenario used by SKB are presented in SKBdoc
1219727, ver 2.0. (Confidential information. Available only for the Swedish Radiation Safety
Authority.) The information is based on data provided by the nuclear power plants. It is secretaccording to The Swedish Competition Act (2008:579) and can be shown to SSM representatives
on request.
The majority of the fuel used in the reactors consists of uranium oxide fuel (UOX). From
Oskarshamn, there will be minor amounts of mixed oxide fuel (MOX), the background to thisis given in Section 2.1.2.

2.1.2 Miscellaneous fuels – background and sources
In the early part of the Swedish nuclear power programme some spent nuclear fuel was reprocessed.
In addition to the nuclear power plants accounted for in Section 2.1.1 there has also been a reactor
for heat production and a research reactor in Sweden. There are also small amounts of fuel residues
from investigations at the Studsvik facility. As a result there will be some MOX fuel and other spent
fuels to be deposited in the KBS-3 repository.
Some spent fuel from Ringhals and Barsebäck was sent to La Hague for reprocessing. This spent
fuel was exchanged in 1986 for spent German MOX fuel in equivalent amounts of plutonium. This
fuel is referred to as “Swap MOX”, and is presently stored in Clab. The amounts are included in
the miscellaneous fuels presented in Section 2.2.2. Some spent fuel from Oskarshamn was sent toSellafield for reprocessing. The plutonium resulting from this reprocessing will be used to manufacture MOX fuel to be used in one of the Oskarshamn reactors. This fuel is included in the spent fuelreference scenario presented in Section 2.1.1.
There was a pressurised heavy water reactor (PHRW) in Ågesta in operation during 1963–1974. The
reactor supplied a suburb in Stockholm with district heating but also small amounts of electricity.
The reactor used natural uranium in the first core and in a second core some slightly enricheduranium as fuel and heavy water as coolant. Some fuel from the first core was reprocessed. The
remaining spent fuel is presently stored in Clab and the amounts are included in the miscellaneous
fuels presented in Section 2.2.2.
The spent metal fuel from the research reactor at the Royal Institute of Technology in Stockholm,
which was in operation during 1954–1970, has been sent to Sellafield for reprocessing. The plutonium from the reprocessing will be used to manufacture MOX fuel that will be used in the BWR
reactors at Oskarshamn. A minor amount of the metal fuel is oxidised and will be disposed together
with the fuel residues from fuel investigations in Studsvik.
Investigations of spent fuel performed at the Studsvik laboratory result in small amounts of fuelresidues. The fuel residues are fixed in different matrices such as epoxy resin, glass, brass and thenput in boxes with the dimensions of a PWR fuel assembly.

2.2

Fuel quantities, burnup and age of the spent nuclear fuel2.2.1

Spent fuel from the reference scenario for operation of the nuclear
power plants
The presented fuel quantities and burnup are based on the fuel stored in the interim storage
facility Clab and the reference scenario for the future operation of the power plants presented in
Section 2.1.1. The total quantity of spent fuel will depend on the energy output of the reactors, which
in turn depends on the operating time and power of the reactors. The energy output is provided by
the fuel assemblies in the reactor core. For a given energy output the exchange rate, and number offuel assemblies, can be kept low if the enrichment and burnup is increased. If assemblies with lower
enrichment and burnup are used the number of fuel assemblies will increase.

16

TR-10-13

The number of fuel assemblies discharged from the Swedish nuclear power plants stored in theinterim storage facility at the end of 2007, and the estimated total number of assemblies for thereference scenario used by SKB are given in Table 2‑2 together with the corresponding estimatedamount of uranium or heavy metal expressed as tonnes initial weight. As previously mentioned the
major part of the spent fuel assemblies from the nuclear power plants will consist of UOX fuel. Only
one of the reactors in Oskarshamn will use a minor amount of MOX fuel assemblies. The BWR
assemblies accounted for in Table 2‑2 also include rod cassettes which consist of fuel rods that havebeen dismantled for various investigations and placed in fuel rod cassettes. Approximately one outof four of the PWR assemblies will contain a control rods cluster. A small amount of the assemblies,
both BWR and PWR, will contain inserts such as start-up neutron sources, boron glass rods and plugs.
Since the beginning of 1970 the burnup of the nuclear fuel has increased from approximately
23 MWd/kgU up to 53 MWd/kgU. The average burnup of the spent nuclear fuel stored in
the interim storage facility is about 34 and 41 MWd/kgU for BWR and PWR fuel assemblies
respectively (December 2007). For the remaining operation, the burnup will increase as a result of
increased power and optimisation of the operation of the reactors. If the enrichment is increased theburnup will also be increased. Alternatively, if the enrichment is not increased, the fuel assemblies
must be switched more often to achieve the same energy output. The resulting average burnup forthe reference scenario is 40.4 and 44.8 MWd/kgU for BWR and PWR fuel assemblies respectively/­SKBdoc 1221579/. The burnup distribution for the BWR and PWR spent fuel assemblies stored in
the interim storage facility, Clab (31 December 2007) and the prognosis for the reference scenario isshown in Figure 2‑1.
To provide an overview of the distribution among the total quantities of BWR and PWR spent fuelto be finally deposited in the KBS-3 repository the information from Figure 2‑1 is expressed in
Figure 2‑2 as tonnes of heavy metal.
Throughout the operation time of the nuclear reactors fuels from different suppliers have been used
and the operation efficiency improved. As a result, there are a number of specific fuel types withdifferent enrichment, burnup and detailed design. A list of different kinds of fuel used in Swedish
reactors is provided in Appendix A.
In Figure 2‑3 the BWR assemblies stored in Clab at the end of 2007 and the assemblies included in
the reference scenario and their burnup and age are plotted for 2045, i.e. the last year of operationof the last reactor to close down. In Figure 2‑4 the corresponding information about the PWR
assemblies is shown. For the assemblies included in the reference scenario large red dots represent
the batch average discharge burnup. The smaller red dots represent the assumed standard deviationin burnup, i.e. ±3 MWd/kgU, averaged over single fuel assemblies included in a batch. Each dot
represents several assemblies. Low burnup assemblies in the reference scenario are from the lastyear of operation of the nuclear power plants.
Table 2‑2. Spent fuel from operation of the nuclear power plants stored in the interim storage
facility, Clab, and total amounts estimated for the SKB reference scenario. The information about
fuel stored in Clab is based on Clab’s safeguards accountancy system. The total number isbased on the reference scenario presented in Section 2.1.1.
Fuel type

Number ininterim storage
31 December 2007

Total number for
SKB reference scenario/­SKBdoc 1221567/

Total initial weight for
the reference scenario(tonnes of U or HM)

BWR assemblies from operation of
the NPP at Barsebäck, Oskarshamn,
Forsmark and Ringhals

21,1941,2

47,4983,4

8,3125PWR assemblies from operation of
the NPP at Ringhals

2,552

6,016

2,7916

The fuel channels, see Figure 2‑5, have been removed from 1,520 of the BWR assemblies.
Including 3 BWR MOX assemblies stored at O1.
3
Including 83 BWR MOX assemblies from the reactors in Oskarshamn.
4
Including rod cassettes, i.e. dismounted fuel rods placed in fuel rod cassettes.
5
Assumed 175 kg U/HM per assembly.
6
Assumed 464 kg U per assembly.
1

2

TR-10-13

17

Number of BWR fuel assemblies

16000
14000

BWR Prognosis

12000

BWR Produced

10000
8000
6000
4000
2000
0
<10

10–20

20–30

30–40

40–50

50–60

Number of PWR fuel assemblies

Burnup intervals (MWd/kgHM)
3500
3000

PWR Prognosis

2500

PWR Produced

2000
1500
1000
500
0
<10

10–20

20–30

30–40

40–50

50–60

Burnup intervals (MWd/kgHM)

Figure 2‑1. Burnup distribution of the spent fuel assemblies stored in Clab 31 December 2007 (greycolours) and prognosis of the burnup distribution for the assemblies resulting from the future operationof the nuclear power plants (red colours) as assumed for the reference scenario used by SKB.

4500
4000

Tonnes HM

3500
3000

PWR Prognosis

2500

BWR Prognosis

2000

PWR Produced

1500

BWR Produced

1000
500
0
<10

10–20

20–30

30–40

40–50

50–60

Burnup intervals (MWd/kgHM)

Figure 2‑2. Average burnup distribution per tonne heavy metal of the BWR and PWR fuel stored in Clab
31 December 2007 (grey colours) and prognosis of the burnup distribution for the future operation of thenuclear power plants (red colours) as assumed for the reference scenario used by SKB.

18

TR-10-13

70

Burnup (MWd/kgHM)

60
50
40
30
20
10
0
0

10

20

30

40

50

60

70

80

Age of BWR assemblies (years)

Figure 2‑3. The burnup and age in 2045 of the BWR assemblies included in the SKB reference scenario.
Assemblies that are currently stored in Clab are marked with black and assemblies from the future operation are marked with red. Large red dots represent batch average discharge burnup and small red dots
represent standard deviation in burnup, i.e. ±3 MWd/kgU, averaged over single fuel assemblies included
in a batch.
70

Burnup (MWd/kgHM)

60
50
40
30
20
10
0
0

10

20

30

40

50

60

70

80

Age of PWR assemblies (years)

Figure 2‑4. The burnup and age in 2045 of the PWR assemblies included in the reference scenario.
Assemblies that are currently stored in Clab are marked with black and assemblies from the future operation are marked with red. Large red dots represent batch average discharge burnup and small red dots
represent standard deviation in burnup, i.e. ±3 MWd/kgU, averaged over single fuel assemblies included
in a batch.

2.2.2 Miscellaneous fuels
The amount of miscellaneous fuels and, if applicable, the number of fuel assemblies in the interim
storage facility as well as the estimated total amount of uranium or heavy metal, expressed as tonnes
initial weight, is presented in Table 2‑3.

TR-10-13

19

Table 2‑3. Miscellaneous fuels stored in the interim storage facility, Clab, and total amounts
estimated for the SKB reference scenario. Information about fuel stored in Clab is based on
Clab’s safeguards accountancy system.
Fuel type

Number ininterim storage
31 December 2007

Total number for the
SKB reference scenario

Total initial weight for
the reference scenario(tonnes of U or HM)

Fuel assemblies from Ågesta

222(1 unirradiated)

222(1 unirradiated)

20
14.1

Swap MOX assemblies (BWR)

184

184

Swap MOX assemblies (PWR)

33

33

8.4

Fuel residues in special boxes from
Studsvik

19

Approximately 25

3Damaged fuel in protection boxes

  0

–

–

The enrichments and burnup of the minor quantities of miscellaneous spent fuel to be deposited canbe summarised as follows.
•Ågesta PHWR fuelThe majority of the fuel assemblies (about 70%) contain natural uranium and the remainder areenriched to 1.35%, except one assembly that contains 2.2% U-235. The burnup ranges from 0 to
10 MWd/kg U.

•

Swap MOX fuel
The Swap MOX fuel has a higher transuranic content than UOX fuel. The average burnup is
16 MWd/kg HM for the BWR assemblies and 31 MWd/kg HM for the PWR assemblies.

•

Fuel residues from Studsvik
The fuel residues from the performed investigations at Studsvik are emplaced in special boxes
with 125 kg U per box. The box has the same dimensions as a PWR fuel assembly. The enrichment varies from low up to about 20% with an average of about 3% U-235 in each box.

•

Damaged fuel
Fuel assemblies damaged so that it is possible that material may fall off are emplaced in protection boxes during storage. Currently there is no such fuel stored in Clab. However, there areassemblies with leaking rods in Clab. Further, some damaged fuel assemblies are currently stored
at the nuclear power plants.

2.3

Fuel parameters of importance for repository design andlong-term safety

As discussed in the Repository production report, Section 2.3, there are some properties of the
spent fuel that are of importance for the design of the engineered barriers and the layout of therepository. There are also some properties that need to be known or predicted with respect to theanalyses performed in the safety assessment. The design of the KBS-3 repository has in turn resulted
in requirements on the selection of fuel assemblies for encapsulation in each canister and on thehandling of the spent fuel before encapsulation.
In this section, the fuel parameters that need to be documented with respect to the design of the final
repository or for the analyses to be performed in the long-term safety assessment are presented.
Where relevant, references are given to the sections in which requirements on the handling arepresented.
In addition to the parameters documented in the handling, the very low solubility of the fuel matrix
is of importance for the long-term safety. The spent fuels from the approved Swedish nuclear power
programme as well as the various miscellaneous fuels are in the oxide form – UOX or MOX. The
fuel matrix in such fuels has very low solubility in a KBS-3 repository environment. This is an
essential property in some scenarios of the long-term safety assessment.

20

TR-10-13

2.3.1

Enrichment, burnup, irradiation and power history and age of the spent
fuel assemblies
The radionuclide inventory is an essential input to the dose calculations in the assessment of thelong-term safety. The burnup data forms the basis for calculations of the radioactivity and radionuclide inventory of the fuel assemblies. The radionuclide inventory and its distribution in the spent
fuel matrix are also affected by the irradiation and power histories of the assemblies. The irradiation
history will affect the amount of activation products formed in the construction materials of thefuel assemblies as a result of the neutron radiation in the reactor. The power history, which is theheat power developed per unit length of fuel rod or fuel assembly over the irradiation period in thereactor vessel, is strongly correlated to the fission gas release (FGR). The FGR reflects the part of
the radionuclide inventory located at the fuel grain boundaries, fractures in the fuel pellets and othergaps within the fuel cladding. The fuel parameters of importance for the radionuclide inventory andits distribution in the fuel matrix are further discussed in Section 6.2.1.
The radioactivity is the source of the heat generation in the spent fuel. The burnup and age of the
spent fuel assemblies will determine the activity content and, thus, the decay power of the spent
fuel. The design of the repository and the assessment of the evolution of the barriers are based on a
maximum decay power in each canister at the time of emplacement.
The radioactivity is also the source of radiation from the spent fuel. The radiation on the canister
surface can cause formation of nitric acid from moist air and corrosive species from radiolysis ofwater. Both these processes can increase the corrosion of the copper canister. The radiation will
depend on the burnup and age of the spent fuel assemblies.
Criticality depends on the amount and geometrical configuration of the fissile material and thesubstances and materials surrounding it. The amount of fissile material in the spent fuel depends onthe fuel type, the enrichment and burnup.
Requirements on the repository design and handling with respect to enrichment, burnup and age ofthe fuel assemblies are presented in Sections 3.1.1 and 3.1.2.

2.3.2 Dimensions and materials
The dimensions of the fuel assemblies will affect the dimensions of the canister. The BWR fuelassemblies contain about 60 up to 100 fuel rods. The fuel rods consist of zirconium alloy tubes filledwith cylindrical fuel pellets. The rods are arranged in square arrays enclosed in a fuel channel. The
cross-sectional area of the fuel assemblies is about 0.141×0.141 m2 and the total length can be up to
about 4.4 m. The PWR fuel assemblies contain 204 or 264 fuel rods, arranged in square arrays. The
cross-sectional area is about 0.214×0.214 m2 and the total length is about 4.3 m.
A BWR and a PWR assembly are illustrated in Figure 2‑5. As previously mentioned, the detailed
designs of the fuel assemblies have been altered as a result of optimisation of the operation of thenuclear power plants and utilisation of the uranium. Detailed information on dimensions for different
BWR and PWR fuel types can be found in Appendix A. During the irradiation in the reactor thedimensions of the assemblies may be altered so that they deviate from the specified.
Requirements on the design of the canister and on the handling of the fuel assemblies with respect totheir dimensions are given in Section 3.1.3.
The material compositions of the fuel assemblies are used as premises in the analysis of the longterm safety. An overview of the materials in typical BWR and PWR fuel assemblies is given in
Table 2‑4. More detailed information is provided in Appendix B.

TR-10-13

21

III

E

VI

V

D

IV

C

A

A
B
C
D
E

I

Length ~4.4 m
Maximum cross section area
141×141 mm
Fuel channel
Fuel rod
Fuel pellet

I
II
III
IV
V
VI

Length ~4.3 m
Maximum cross section area
214×214 mm
Control rod cluster
Guide tube for control rod
Fuel rod
Fuel pellet

II

B

Figure 2‑5. Arbitrary BWR (left) and PWR (right) assemblies. Data based on specifications provided by
the nuclear fuel suppliers.
Table 2‑4. Overview of the materials in typical BWR and PWR fuel assemblies. Data based on
specifications provided by the nuclear fuel suppliers. Control rod clusters in PWR assemblies
are not included (for details, see Appendix B).
Material composition

BWR
PWR
Svea 96 Optima 2
Areva 17×17
Weight in 1 fuel assembly (kg)

Fuel
U-tot

175

464

O

23

62

Zirconium alloys

49

108

Stainless steel

  –

  3

Zirconium alloys

32

  –

Stainless steel

  8

  –

Stainless steel

  5

12

Zirconium alloys

  –

21

Nickel alloys

  1

  2

Cladding material

Fuel channel

Other constructions (bottom and top plate, spacers etc)

Four elements of particular interest for the assessment of the long-term safety are nitrogen (N),chlorine (Cl), nickel (Ni) and niobium (Nb). The contents of these elements in the construction
materials for a PWR Areva 17×17 assembly and a BWR Svea 96 Optima 2 assembly are given in
Table 2‑5. The content of these elements is based on the material compositions given in Appendix B.
The variability of these elements in the different kinds of BWR and PWR assemblies to be deposited
has been investigated by randomly selecting a number of fuel types and comparing the amounts ofconstruction materials in these assemblies with the amounts in Svea 96 Optima 2 and Areva 17×17
respectively. The conclusion is that the amounts will be similar for all BWR and PWR assemblies
respectively.

22

TR-10-13

Table 2‑5. The content of N, Cl, Ni and Nb in the construction material for typical BWR and PWR
fuel assemblies.
Element

N

Weight in 1 fuel assembly (kg)
BWR Svea 96 Optima 2
Cladding Other
Fuel channel

Total

PWR Areva 17×17
Cladding
Other

Total

2·10–3

Cl
Ni
Nb

2·10

–2

2·10–3

5·10–3

9·10–35.5·10–3

5.8·10–3

1.1·10–2

6·10–68·10–61.4·10–5

3·10–61.4·10–5

1.7·10–5

1.1

0.86

1.99

0.27

2.19

2.46

8.2·10–3

8·10–4

9·10–3

1.08

9·10–2

1.17

There are no requirements on the handling with respect to the material composition of the fuelassemblies.

2.3.3 Encapsulated liquids and gases
Liquids and gases that remain in the canister when it is sealed may cause corrosion of the cast ironinsert. Water will lead to anaerobic corrosion of the cast iron insert. Radiolysis will form nitric acidfrom water and air in the canister. Nitric acid will corrode the insert and may also cause stress corrosion cracking in areas with tensile stresses.
Requirements on the handling of the spent nuclear fuel with respect to liquids and gases arepresented in Section 3.1.4 Encapsulated liquids and gases.

TR-10-1323

3Requirements on the handling of the spent fuelIn this chapter, the requirements on the handling of the spent fuel related to its final disposal in theKBS‑3 repository are presented. The chapter also contains design premises imposed on the canisterby the spent nuclear fuel. The presentation has been divided into two sections:••requirements related to repository design and long-term safety,
requirements related to the operation of the KBS-3 system.

3.1

Requirements related to repository design andlong‑term safety

3.1.1 Decay power
The decay power of a fuel assembly depends on its burnup, age and the mass of heavy metal (HM).
The decay power of the spent nuclear fuel will influence the temperature increase in the final repository. High temperatures will impact the properties of the engineered barriers and may affect theirbarrier functions. High temperatures will also generate rock stresses that may cause rock fracturing.
As a consequence of this, the KBS-3 repository must be designed so that a maximum allowed
temperature will not be exceeded.
In Design premises long-term safety the following design premise is stated for the maximum
allowed temperature.
•

The buffer geometry (e.g. void spaces), water content and distances between deposition holes
should be selected such that the temperature in the buffer is <100°C.

In order to determine a repository design where the temperature stays below the maximum allowedin the buffer, the total decay power in the canisters must be known. The total decay power in a canister will depend on the burnup and age of the fuel assemblies and also on the number of assemblies in
the canister. The following requirement and criterion are set for the selection of fuel assemblies forencapsulation.
•

Requirement on handling: The fuel assemblies to be encapsulated in any single canister shallbe selected with respect to burnup and age so that the total decay power in the canister will not
result in temperatures exceeding the maximum allowed in the buffer.

•

Criterion: The total decay power in each canister must not exceed 1,700 W.

The background to these requirements is discussed in the Repository production report,
Section 2.3. Note that the allowed decay power may be altered as a result of optimisation when moredetailed information on burnup, operation times and conditions at repository depth are available. It
may also be possible to allow higher decay power in peripheral deposition holes.

3.1.2 Criticality
Criticality must always be prevented outside the reactor vessel, and the following design requirement
is set for the canister; see Design premises long-term safety, Section 3.1.4.
•

The spent fuel properties and geometrical arrangement in the canister should be such thatcriticality is avoided even if water should enter the canister.

In the analysis of the propensity or potential for criticality of the fuel assemblies placed in thecanister, the sensitivity of canister material composition and dimensions are investigated, see
Section 4.4.1.

TR-10-1325

The sensitivity analyses are based on the assumption that the insert is made of nodular cast ironwith an iron content of at least 90%. The iron in the insert acts as a neutron reflector. Alloying elements occurring in nodular iron that are more potent neutron reflectors than iron are silicon (Si) andcarbon (C). In the analysis it was concluded that the content of these substances shall be kept below
6% (Si) and 4.5% (C) in order not to increase the propensity for criticality. Further, the propensity
for criticality increases if the assemblies are placed close together. The loading curves presented in
Section 4.4.1 are based on the closest possible distances based on the acceptable distances between
the channel tubes for the reference design of the canister.
The following requirement and criterion are set for the selection of fuel assemblies to be encapsulated.
•

Requirement on handling: The fuel assemblies to be encapsulated shall be selected with respect
to enrichment, burnup, geometrical configuration and materials in the canister so that criticality
will not occur during the handling and storage, even if the canister is filled with water.

•

Criterion: The effective multiplication factor (keff) must not exceed 0.95 including uncertainties.

3.1.3 Dimensions and spacing devices
The dimensions of the BWR and PWR fuel assemblies, including alterations that may occur as aresult of the irradiation in the nuclear reactor, shall be considered in the design of the canister inserts.
Two types of canister inserts with the same length and diameter provided with channel tubes withdifferent inner dimensions to accommodate BWR and PWR fuel assemblies, respectively, will be
manufactured; see the Canister production report.
SKB has decided that it shall be possible to encapsulate all spent fuel from the Swedish nuclear
power programme, i.e. also the Ågesta fuel, the swap MOX fuel, the Studsvik fuel residues andthe special boxes containing fuel rods, in either BWR or PWR canisters, and the following design
requirement is set for the canister.
•

Design requirement: The dimensions of the fuel channel tubes of the insert shall be adapted to thedimensions of the spent fuel to be deposited.

•

Design premises: The length of the longest BWR or PWR assembly, including induced length
increase. The cross section of the largest BWR and PWR fuel assemblies, including deviations
due to deformations during operation.

The measures that shall be used in the design of the channel tubes of the insert are given in
Table 3‑1.
Table 3‑1. Design measures for the fuel channel tubes of the insert.
DetailBWRLongest assembly

4,441 mm

Before irradiation.

Induced length increase

14 mmWhen determining the length of the longest
assembly the length before irradiation and theinduced length increase is considered.

Largest cross section

141×141 mm

214×214 mm

Before irradiation.

Deviations due todeformations during
operation

145.5×145.5 mm

228×228 mm

Cross sections of BWR transport cask, and PWR
storage canister respectively. All assemblies
in Clab have been placed in these casks orcanisters, i.e. these cross sections are sufficientwith respect to occurring deviations due todeformations during operation.

26

PWR

Comment

TR-10-13

Since the dimensions of the miscellaneous fuel types deviate from those of the BWR and PWR
assemblies, boxes or spacing devices are required in order to keep them in position in the channel
tubes of the canister. The different kinds of BWR and PWR assemblies may also vary in dimensions
and require spacing devices. Based on this, the following requirements are put on the handling of the
spent fuel.
•Requirement on handling: The miscellaneous fuels shall be encapsulated in either BWR or PWR
canisters.

•

Requirement on handling: Devices that keep the miscellaneous fuel types in position shall beplaced in the channel tubes of the canister.

•

Requirement on handling: Devices that keep BWR and PWR assemblies in position shall, if
necessary, be placed in the channel tubes of the canister.

Detailed design premises for the spacing devices will be provided at a later stage.

3.1.4 Encapsulated liquids and gases
Nitric acid formed from radiolysis of water and air remaining in the canister when it is sealed may
cause corrosion of the cast iron insert and the copper shell. The fuel assemblies are stored in water
before encapsulation and in the case of a cladding leak there may be water inside the fuel rods.
Consequently, in order to limit the amount of water and air, the fuel assemblies must be dried before
encapsulation and the atmosphere inside the canister changed and inspected.
With respect to this the following design premise is stated for the canister; see Design premises
long-term safety Section 3.1.5.
•

Design premise: The amount of nitric acid formed within the insert is limited by changing theatmosphere in the insert from air to > 90% argon. The maximum amount of water left in theinsert is set to 600 g.

This is the background to the following requirements on the handling of the spent nuclear fuel.
•Requirement on handling: Before the fuel assemblies are placed in the canister they shall bedried so that it can be justified that the allowed amount of water stated as a design premise forthe canister is not exceeded.

•

Criterion: The amount of water left in anyone canister shall be less than 600 g.

•

Requirement on handling: Before the canister is finally sealed, the atmosphere in the insert shallbe changed so that acceptable chemical conditions can be ensured.

•

Criterion: The atmosphere in canister insert shall consist of at least 90% argon.

Equipment for drying of fuel assemblies and exchange of atmosphere shall be designed to conformto these requirements and criteria. These requirements and criteria shall also be considered in the
instructions for the handling of the fuel in the encapsulation plant. The required inspections arefurther discussed in Chapter 4.

3.1.5 Radiation
The radiation at the canister surface may result in the formation of nitric acid and other corrosive
species that may cause increased corrosion of the copper canister surface in the final repository. If
the radiation is limited, these processes can be neglected; see Design premises long-term safety,
Section 3.1.5. The radiation at the surface of the canister will depend on the radiation shielding
provided by the insert and copper shell and the radioactivity of the encapsulated spent fuel. The
radioactivity of the spent fuel will depend on the burnup and the age of the fuel assemblies. The
burnup and age of the encapsulated spent fuel will, consequently for a specific design of the canister,
set a limit for the radiation dose rate at the canister surface. Based on this, the following requirement
and criterion is set for the handling of the spent fuel.

TR-10-1327

•Requirement on handling: It shall be verified that the radiation dose rate on the canister surface
will not exceed the level used as a premise in the assessment of the long-term safety.

•

Criterion: The radiation dose rate at the surface of the canister must not exceed 1 Gy/h.

This criterion shall be verified for the canister after the placement of the fuel assemblies selected for
encapsulation; see Sections 4.4.1 and 4.7.2.

3.2

Requirements related to the operation of the KBS-3 system

3.2.1 Encapsulation
The encapsulation of the spent nuclear fuel shall be adapted to the nuclear power programme so thatthe costs and environmental impact are minimised. With respect to this, provided that the selectedassemblies conform to the acceptance criteria for decay power and criticality, it is an advantage if thecanisters are filled to their maximum capacity and the following objective is set for the selection ofassemblies.
•

Requirement on handling: The number of canisters shall be minimised and, if possible, allassembly positions in the deposited canisters shall be filled.

At the end of the nuclear power programme, fuel assemblies with high burnup are expected. Unlessthese assemblies can be combined with assemblies with low decay power, they require long interim
storage times to conform to the decay power criterion if all positions in the canister are to be filled.
This means that canisters that are not filled to their maximum capacity may have to be deposited
to conform to the decay power criterion. The number of such canisters will, in addition to burnup,
depend on when the encapsulation is initiated, on the length of the operation period and on theencapsulation rate. The facilities shall be designed for an encapsulation and deposition rate of up
to 200 canisters per year.
The fuel assemblies are stored in pools with water, inside storage canisters with 16 or 25 BWR
assemblies and 5 or 9 PWR assemblies in each storage canister. It is an advantage, regarding safety
and radiation protection and costs, if the number of movements of the fuel assemblies can be
minimised. Based on this, the following objectives are set for the selection of assemblies.
•

Requirement on handling: The number of lifts and movements of fuel assemblies shall beminimised.

•

Requirement on handling: If possible, storage canisters shall be emptied before they are broughtback from the encapsulation plant to the interim storage facility.

3.2.2 Operational safety and radiation protection
For the operational safety assessments of the facilities and radiation protection considerations of thetransports within the KBS-3 system, the radioactivity content in the canister and the radiation at thecanister surface must be known. Based on this, the following requirement is set for the handling.
•

Requirement on handling: It shall be verified that the radioactivity content and radiation level on
the canister surface will not exceed the contents and levels used as premises in the assessments of
the operational safety.

The criteria for conformity to this requirement as well as the verification are addressed in the safety
report for the Clink facility, SR‑Operation and the safety report for the transports of encapsulated
spent fuel.

3.2.3 Control of nuclear material – Safeguards
Sweden’s commitments regarding non-proliferation and the planned actions to control the management of fissile material within the KBS-3 system is presented in /­SKBdoc 1172138/. After the spent
fuel has been encapsulated it is no longer possible to control individual assemblies and each canister
will constitute a unit for the account and control of nuclear material. The requirement on marking of
the canister that follows from this is described in the Canister production report, Section 2.3.2.
28

TR-10-13

4

The handling of the spent nuclear fuel4.1

Overview

The handling of the spent nuclear fuel comprises the following main parts:•••

transport and delivery of fuel assemblies,
interim storage,
selection of assemblies and encapsulation.

A flow chart for the spent fuel line and its stages, with references to the sections in which the different stages are presented, is given in Figure 4‑1. The interim storage and encapsulation of the spent
nuclear fuel will take place in Clink (Central interim storage and encapsulation plant). The Clink
facility will consist of the current interim storage facility (Clab) and the encapsulation plant to beconstructed in connection to Clab.

4.2

Transport and delivery of fuel assemblies

4.2.1 Activities
The spent nuclear fuel is transported to the interim storage facility in accordance with applicableregulations for transports of radioactive material.
When the transport arrives at Clab (later Clink), the transport cask is inspected and transported
to a cell for cooling. When the temperature has stabilised, the transport cask is transported to thecask pool and connected to the unloading pool. The fuel assemblies are unloaded and placed in thestorage canisters in which they are kept during the interim storage. If the assemblies are damaged,
or suspected to be leaking, the storage canister is transported to a service pool and the assemblies aresearched for damages. If necessary, damaged assemblies are put in protective boxes and placed in
special storage canisters.
Each storage canister and each fuel assembly is marked with a unique identity code. When the fuelassembly is put in the storage canister, its identity is linked to the identity of the storage canister andits position in the storage canister is registered.

Transport and deliveryof fuel assemblies

Interim storage

Selection of fuelassemblies andencapsulation

Transport and delivery offuel assemblies
Section 4.2

Interim storage
Section 4.3

Selection of assemblies
Section 4.4
Delivery to encapsulationbuilding
Section 4.5
Drying of fuel assemblies
Section 4.6
Placement in canister
Section 4.7
Replacement of atmospherein the canister insert
Section 4.8

Figure 4‑1. Flow chart for the spent fuel line and its stages. A flow chart for the handling of the encapsulated spent fuel is provided in the Canister production report, Figure 6-1.

TR-10-1329

4.2.2 Inspections
Before a transport is approved the nuclear power plants announce which fuel assemblies they intend
to send and the properties of the assemblies. SKB inspects the documentation to ensure that theassemblies are allowed for transportation by the transportation system and for interim storage in
Clab (later Clink). Then the transport can be performed.
When the fuel assemblies are unloaded, their identities are visually inspected. The inspection is
performed with a camera and the information is documented. Properties of importance for thecontrol of nuclear material, safeguards, are stored in a special database for this purpose. Propertiesof importance for the operation of SKB’s facilities and long-term safety are stored in other databases
containing all relevant information about the fuel assemblies. Regarding the burnup SKB intends
to use data provided by the nuclear power plants. The burnup must be regularly calculated andaccurately measured to achieve a reliable and efficient operation of the reactors. The registration of
burnup during operation is required by the Swedish Radiation safety authority (SSM) and the data isquality assured. If required, the burnup can also be measured by γ‑scanning.

4.3

Interim storage

4.3.1 Activities
When the placement and registration of assemblies in the storage canister is finished, the storage
canister is transported to a fixed position with a unique identity in one of the storage pools. The
assemblies are stored until they are selected for encapsulation. They are selected based on theirdecay power; see further Section 4.4.1.
The storage is constantly monitored and all movements of the storage canisters are registered.

4.3.2 Inspections
If the storage canisters or fuel assemblies are moved, their identities are inspected and their new
positions registered. No other inspections of the fuel are required during the interim storage. If
required, it is possible to measure the radiation and decay power of the assemblies in the interim
storage facility. The operation of the facility, e.g. levels and temperatures of the water in the storage
pools, is regularly measured.

4.4

Selection of assemblies

4.4.1 Activities
The selection of fuel assemblies for encapsulation is iteratively performed by applying the requirements presented in Section 3.1 in the following order.
•

The fuel assemblies to be encapsulated shall be selected with respect to burnup and age so thatthe total decay power in the canister will not result in temperatures exceeding the maximum
allowed. The total decay power in each canister must not exceed 1,700 W.

•

The number of canisters shall be minimised and, if possible, all assembly positions in thedeposited canisters shall be filled.

•

The fuel assemblies to be encapsulated shall be selected with respect to enrichment and burnup
and the geometrical configuration and materials in the canister so that criticality will not occur
during the handling and storage of canisters even if the canister is filled with water. The effective
multiplication factor (keff) must not exceed 0.95 including uncertainties.

•

It shall be verified that the radiation dose rate on the canister surface will not exceed the level
used as a premise in the assessment of the long-term safety. The radiation dose rate at the surfaceof the canister must not exceed 1 Gy/h.

•

The number of lifts and movements of fuel assemblies shall be minimised.

30

TR-10-13

•

If possible, storage canisters shall be emptied before they are brought back from the encapsulation plant to the interim storage facility.

The selection process can be summarised as follows.
1. Compile information for the selection.
2. Preliminarily selection – based on decay power and the objective to fill all assembly positions in
the canisters to be deposited.
3. Check criticality – adjust the selection in case of non-conformity to the criterion for criticality.
4. Check radiation dose rate on the canister surface – adjust the selection in case of non-conformity
to the criterion for maximum allowed radiation dose rate.
5. Lifts and movements – investigate the number of lifts and movements of assemblies and storage
canisters and adjust the selection if the number of lifts can be reduced and the selection still
conforms to criteria for decay power and criticality.
6. Final selection – determine a selection and make a plan for transport of storage canisters andassemblies.
Compile information for the selection

The preliminary selection will be based on documented and calculated information of the total
inventory of assemblies in interim storage at the time of encapsulation. In order to optimise theselection the full inventory of assemblies, i.e. both the assemblies currently in interim storage andthe assemblies from the future operation, are considered in the strategy for selection. Before theselection is made, the decay power is calculated for all assemblies and loading curves to check thecriticality criterion are produced.
The decay power for each assembly is calculated from its individual burnup data and the date theassembly was taken out from the reactor core. All data are provided by the nuclear power plants.
A well documented and verified code will be used for the calculations. In Figure 4‑2 the decay
power as a function of age is illustrated for BWR assemblies with different burnup.
500

30 MWd/kg HM
38 MWd/kg HM
45 MWd/kg HM
50 MWd/kg HM
55 MWd/kg HM
60 MWd/kg HM

450

Decay power (W)

400
350
300
250
200
150
100
50
0
10

20

30

40

50

60

70

80

90

100 110 120 130 140 150

Age of spent fuel assembly (years)

Figure 4‑2. The decay power as a function of age for BWR assemblies with different burnup. (The allowed
average decay power per encapsulated assembly is 142 W for a canister with 12 identical assemblies.)/­SKBdoc 1179234/.

TR-10-1331

Criticality is checked by calculating loading curves. In those calculations, combinations of enrichments and burnups that will result in a multiplication factor (keff) of 0.95 for the encapsulated
assemblies are derived /­SKBdoc 1193244/. The calculations are based on typical BWR and PWR
assemblies, i.e. Svea Optima 2 and Areva 17×17, respectively, and on the reference design of thecanister and typical properties of bentonite and rock. Further, it is assumed that identical assemblies
occupy all positions in the canister. In the calculations, a systematic investigation of uncertainties ismade. All parameters with potential impact on the criticality are investigated. Parameters that can beshown to be insignificant are set to a typical value while parameters that are significant are set so as
they favour criticality.
The calculated loading curves and the combinations of average burnup and enrichment for theassemblies that currently are stored in Clab are given in Figure 4‑3 and Figure 4‑4. Fuel assemblies
with a combination burnup/enrichment that are plotted above the loading curves in Figure 4‑3 andFigure 4‑4 will have a keff that exceeds 0.95 and will thus not conform to the criterion for criticality.

Initial enrichment (% U-235)

5

4

3

2

Fuel assemblies
Loading curve-actinides
Loading curve-actinides+fission products

1

0

0

10

20

30

40

50

60

Burnup (MWd/kgU)

Figure 4‑3. Loading curves for BWR-canisters with 12 identical fuel assemblies and enrichment andaverage burnup for the assemblies currently stored in Clab /­SKBdoc 1193244/.

Initial enrichment (%U235)

5

4

3

2

1

Fuel assemblies
Loading curve-actinides
Loading curve-actinides+fission products

0
0

10

20

30

40

50

60

Burnup (MWd/kgU)

Figure 4‑4. Loading curves for PWR-canisters with 4 identical fuel assemblies and enrichment andaverage burnup for the assemblies currently stored in Clab /­SKBdoc 1193244/.

32

TR-10-13

As can be seen from Figure 4‑3 and Figure 4‑4, respectively, un-irradiated BWR assemblies withenrichment higher than 3.5% and un-irradiated PWR assemblies with enrichment higher than
2.4% will be plotted above the loading curves. The enrichment for the fuel assemblies currently
stored in Clab is typically between 3.6–4.2% U-235, and enrichment up to 5% U-235 is allowed.
Consequently, as can be seen from the loading curves, credit must be taken for burnup in order toconform to the criticality criterion.
Preliminary selection

When the decay power is calculated for all assemblies, the preliminary selection of assemblies to be
encapsulated is made. With the exception of the fuel residues in protection boxes from Studsvik (seeTable 2‑3), the preliminary selection is made from all the BWR and PWR assemblies actually stored
in the Clink facility at the specific time, and from the miscellaneous fuels (Section 2.2.2).
The fuel residues from Studsvik are encapsulated separately since some of them contain epoxy resinthat may build up pressure from gases generated by radiolysis and must not be placed together withassemblies with relatively high decay power and radioactivity.
With respect to the objective to fill all assembly positions in the canisters, the total decay power shalllie as close as possible to the maximum allowed. The assemblies are selected so that the conformity
to the decay power criterion is ensured. Currently, the selection is made so that the sum of thecalculated decay powers for the selected assemblies is 1,650 W or less. The limit of 1,650 W is set tocover divergences between the calculated and actual decay power so that the actual decay power will
always be below 1,700 W. The divergences between calculated and actual decay power have basedon comparisons between calculated and measured decay powers been estimated to be 2% /­SKB 2006/.
Check of criticality

When the selection based on the decay power has been made it is checked against the criterion to
avoid criticality. If keff < 0.95 for all individual assemblies, i.e. if their combination of enrichment
and burnup, lies under the calculated loading curve, they can be encapsulated without further checks.
If there are assemblies with keff > 0.95, the criticality is calculated for the full set of preliminary
selected assemblies. In these calculations, the assembly with keff > 0.95 is placed in the canister inthe worst position for a potential criticality. If the calculations show that keff is above 0.95 for thewhole canister, that selection of assemblies is not encapsulated. Instead a new set of assemblies isselected.
There will be low burnup assemblies with keff > 0.95, that lies on the criticality side of the loading
curve. These assemblies can be combined with high burnup assemblies that lies with equal or larger
distance from the loading curve on the non-criticality side, so that the specific set of assemblies
conform to the criticality criterion, keff < 0.95. If it is not possible to find a set of assemblies that conform to the criticality criterion, the low burnup assemblies can be encapsulated alone in a canister.
Should it neither be possible to combine the low burnup assemblies with high burnup assemblies nor
to encapsulate them alone to conform to the criticality criterion, the ultimate measure is to alter thegeometry, i.e. to reconstruct the assembly.
Check of radiation

The maximum allowed decay power in a canister will, since the decay power is a result of theradioactivity of the spent fuel assemblies, also restrict the radiation at the canister surface. The
radiation dose rate at the canister surface has been calculated for the reference design of the canister/­SKBdoc 1077122/. The results show that highest radiation dose rate to material in contact with thecanister is obtained for PWR canisters. The obtained dose rate is 0.18 Gy/h on limited areas of thecanister tube surface. The burnup and age of the assemblies was set to 60 MWd/kgU and 30 years,
i.e. an overestimation since this combination of assemblies is not allowed with respect to the decay
power criterion. The average radiation dose rate from the copper tube surface is calculated to be
0.055 Gy/h. These results, which give an ample margin to the design premise (1 Gy/h), imply that
all canisters conforming to the criterion for maximum allowed decay power will also conform to thedesign premise for maximum allowed radiation dose rate at the canister surface.

TR-10-1333

Consequently, given the reference design of the canister, Canister production report, Chapter 3,
and the current criteria for decay power and radiation, there is no need to calculate the radiation dose
rate on the surface of individual canisters.
Lifts and movements

When the preliminary selection has been made and the criticality checked, the storage canisters
in which the assemblies are stored are identified. Information on the full set of assemblies in theselected storage canisters is compiled and it is investigated whether several selections for encapsulation can be made from the same storage canisters. If not, it is investigated whether a storage canistercan be changed for another containing a more suitable set of assemblies. If it is anticipated that achange is needed in order to minimise the number of lifts and movements of storage canisters, theselection of assemblies will be made for several canisters at the same time.
Final selection

When all calculations and checks according to the description above have been performed, the selection is documented and a plan for the encapsulation is made.

4.4.2 Inspections
The calculated decay powers and loading curves shall be reviewed in accordance with SKB’s
management system. Input data as well as computer programmes and calculations are reviewed.

4.5

Delivery to the encapsulation building

4.5.1 Activities
When a selection of assemblies has been determined and a plan for the encapsulation made, the storage canisters in which the selected assemblies are stored are lifted from their positions in the storage
pools and transported to the encapsulation building. The transport is performed under water, via thefuel elevator in the interim storage building and the connecting pool in the encapsulation building,to the handling pool in the encapsulation building. Up to 12 storage canisters can be placed in thehandling pool. When all storage canisters with selected assemblies have been moved to the handlingpool, the assemblies to be encapsulated for deposition in a specific canister are reloaded to a transfer
canister used for transports within the encapsulation building. The transfer canister is loaded withthe full set of assemblies to be encapsulated, i.e. either 12 BWR or 4 PWR assemblies (or fewer
in unfilled canisters). The location of the assemblies and altered content in the storage canisters is
registered in the database for control of fissionable material and in SKB’s databases for operation of
the facilities.
Damaged assemblies that are stored in protective boxes are placed with the protective box in thetransfer canister and the lid is removed before the transfer canister is placed in the drying position.

4.5.2 Inspections
The identities of the storage canisters and assemblies in the storage canister are visually inspectedboth before they are lifted from the storage pools and when they are delivered to the handling pool.
The inspection can be documented by a camera. The documented decay power, burnup, enrichment
and other information of importance for the encapsulation is inspected for each assembly.
If there are uncertainties regarding the decay power, it can be measured for verification. This isperformed by measuring the γ-radiation from the assembly. All measurements can be performed in
the handling pool.

34

TR-10-13

4.6

Drying of fuel assemblies

4.6.1 Activities
The drying of assemblies shall conform to the following requirement.
•

Before the fuel assemblies are placed in the canister they shall be dried so that it can be
warranted that the allowed amount of water stated as a design premise for the canister is notexceeded.

When the transfer canister is filled with the assemblies to be encapsulated, it is lifted from thehandling pool and kept hanging so that the water can run and drip off. After that, it is transported
to one of the two drying positions in the handling cell. The drying is performed by decreasing thepressure in the drying cell so that all water is vaporised. The vapour is conducted to a system forcondensation and treatment of process water. Damaged assemblies that are kept in protective boxes
are dried in the box.

4.6.2 Inspections
In Design premises long term safety it is stated: “The maximum amount of water left in the insert
is set to 600 g.” Consequently, it shall be verified that the water content in the canister insert is
600 g or less when it is sealed. The relationship between temperature, pressure and humidity are
well known and by measuring the pressure in the cell it can be verified that the fuel assemblies are
sufficiently dry when they leave the cell. If water remains in the system, the pressure in the cell willrise after the drying is finished.

4.7

Placement in the canister

4.7.1 Activities
The canister in which the selected fuel assemblies will be encapsulated is connected to the handlingcell. Each canister to be deposited shall have a unique identity; this identity is read and registered.
A guiding frame is placed on top of the canister to protect it and guide the assemblies to their correct
position (channel tube). Before the fuel assemblies are lifted from the transfer canister and placed in
the canister, their identity is registered. Fuel assemblies significantly smaller than typical BWR or
PWR assemblies, which may be damaged due to punches when the canister is moved or lifted, areprovided with distance devices that prevent them from moving in the channel tubes in the canister.
This will be necessary for the miscellaneous fuels (Ågesta, Studsvik and swap MOX). The assemblyis then lifted to the channel tube in which it will be placed and lowered into the canister. With respect
to the operational safety, to reduce the radiation on the canister surface, the assemblies generating
more radiation, i.e. MOX assemblies or the assemblies with the largest decay power are placed in
the centre of the insert in BWR canisters. When all assemblies from the transfer canister are placedin the canister, the guiding frame is removed and the identities of the assemblies are registered andassociated to the identity of the canister. The channel tubes in canisters with less than twelve or four
assemblies are left empty. Then the steel lid is put on the insert.
Damaged assemblies that are stored in protective boxes are lifted from the boxes to the canister.
The boxes are then inspected and material that may remain in them is taken care of. If required it
can be placed in a canister and encapsulated. If material should come off from the assembly when
it is handled in the cell this is taken care of, and if required encapsulated. This is described in
Chapter 5, of the safety report for the interim storage and encapsulation plant – Clink.

4.7.2
Inspections
Before the steel lid is placed on the insert, the identities of the assemblies in the canister are
inspected and checked against the planned content, i.e. the selected assemblies. The inspection
is visual and can be documented by photograph.

TR-10-1335

Before the atmosphere in the insert is changed and the canister is sealed the γ-radiation at the canister surface is measured. It shall be verified that the radiation dose rate including any contribution
from neutrons does not exceed 1 Gy/h.

4.8

Replacement of atmosphere in the canister insert

4.8.1 Activities
The atmosphere in the insert is replaced by argon. The position for sealing of the insert has evacuation equipment that is connected to a valve in the steel lid of the canister insert. The pressurein the insert is then lowered by evacuation of the air and argon enters the insert. The atmospherereplacement is then repeated and the composition of the evacuated gas is investigated. If the content
of argon is not above 90%, another repetition of the process is performed. After the atmosphere
has been changed the weld joint surface is checked and the copper lid is placed on the canister.
The canister can then be transferred to the welding station for sealing.
4.8.2 Inspections
The content of argon of the evacuated gas is inspected after each replacement of the atmosphere in
the canister insert. The argon concentration of the evacuated gas shall be greater than 90%.

36

TR-10-13

5

The canisters to be deposited

In this chapter the number of canisters that is required to deposit the spent fuel accounted for in
Chapter 2 is assessed. The assessment is based on the requirements on the handling and selectionof assemblies presented in Chapter 3 and Section 4.4.1 respectively. In addition, the encapsulation
period is postulated.

5.1

Factors that will affect the number of canisters

The number of canisters to be deposited will depend on:•

the number of fuel assemblies in the canisters,•

the number of spent fuel assemblies to be encapsulated,•the burnup of the assemblies,•

the allowed decay power in the canister,•when the encapsulation is initiated and the encapsulation rate.

The maximum number of BWR and PWR assemblies in a canister is twelve and four, respectively.
The number and burnup of the assemblies to be deposited will depend on the thermal powers,
number and type of assemblies in the reactor core and operation times of the nuclear power reactors(see Table 2‑1), and on the past and planned average target burnups.
The allowed total decay power of the assemblies encapsulated in a canister will ultimately dependon the decay power that can be accepted for the temperature in the buffer not to exceed 100°C. The
temperature in the buffer will depend on the thermal properties of the buffer and bedrock and on the
distances between deposition holes; see Underground openings construction report, Section 4.2.1.
The current layout of the final repository is based on a maximum decay power in a canister of
1,700 W.
The decay power of the fuel assemblies will decrease with time; consequently, the age of theassemblies at encapsulation will affect the decay power. The age of the spent fuel assemblies at any
time during the encapsulation period will depend on the operation times of the nuclear power plants
and on when the encapsulation and deposition is initiated and completed.
The assemblies for encapsulation shall be selected with respect to their burnup and age so that they
conform to the criterion for maximum allowed total decay power in a canister.
In Figure 5‑1 the relation between burnup and age of the spent fuel assemblies to reach 1,700 W
in a BWR and PWR canister, respectively, is shown, assuming that all positions in the canister areoccupied by assemblies with the same decay power.
From the interpolation of data in Figure 5‑1, the interim storage time required for the averageburnup BWR and PWR assembly to reach 1,700 W in the canister can be derived. For the spent
fuel from the reference scenario for the operation of the nuclear power plants presented in
Section 2.1.1 this time is 38.7 years for the average burnup BWR assembly and 37.9 years for the
average burnup PWR assembly. If all positions in all canisters are to be filled and their total decay
powers not exceed 1,700 W, the average interim storage time must be at least 38.7 and 37.9 years
for BWR and PWR canisters respectively. Shorter interim storage time will mean that all positions
cannot be filled in all canisters and longer interim storage time will result in lower total decay powerin some canisters.

TR-10-1337

80

Burnup (MWd/kgU)

12 BWR fuel assemblies
4 PWR fuel assemblies

60

40

20

0
20

30

40

50

60

Age of spent fuel assembly

Figure 5‑1. Burnup and age of spent fuel assembly required to reach a decay power of 1,700 W in a BWR
and PWR canister respectively assuming that all positions are occupied by assemblies with the same decay
power.

5.2

Simulation of the encapsulation

The encapsulation of the spent nuclear fuel to be deposited has been simulated assuming that theencapsulation is initiated in 2023 and completed 25 years after the last nuclear power plant is takenout of operation (2070) /­SKBdoc 1221567/. The assemblies were selected so that the conformity
to the decay power criterion was ensured, see Section 4.4.1. Further, the simulation is based on theencapsulation rate given in Table 5‑1.
The assemblies stored in Clab as of December 31th 2007 and the assemblies included in the SKB
reference scenario for the future operation were included in the simulations. With exception of thefuel residues from Studsvik, which require all together seven canisters, the miscellaneous fuels presented in Table 2‑3 were also included in the simulation. The total numbers and kinds of assemblies
included in the simulation are compiled in Table 5‑2. The burnup and age of the BWR and PWR
assemblies included in the simulation are illustrated in Figure 2‑3 and Figure 2‑4. In the simulations
it is anticipated that no more than one MOX assembly will be encapsulated in any canister.
For each simulated year of encapsulation the inventory of assemblies in the Clink facility and theirdecay power was calculated. For assemblies in interim storage, documented data were used and forthe remaining assemblies, the planned reactor operation information was used as input. Other datarequired for the calculations were based on typical BWR and PWR assemblies, i.e. Svea Optima 2
and Areva 17×17, respectively.
Table 5‑1. Encapsulation rate assumed in the simulation of canisters to be deposited in theKBS-3 repository.
Year

Encapsulation rate (canisters/yr)
BWR
PWR
Total

2023

17

6

2024

58

22

80

2025–2027

87

33

120

2028–2054

109

41

150

2055–2069

73

27

100

2070

77

13

90

38

23

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Table 5‑2. Total numbers and kinds of spent fuel included in the simulation of the encapsulation.
Kind of spent fuelNumber ofassemblies

BWR UOX
From the reactors B1, B2, F1, F2, F3, O1, O2, O3 and R1
Including rod cassettes, i.e. dismounted fuel rods placedin fuel rod cassettes.
PWR UOX
From the reactors R2, R3 and R4

47,4156,016Ågesta (encapsulated in BWR canisters)

222

BWR MOX
From Oskarshamn (O1 and O3)
Swap from Germany

83
184

PWR MOX
Swap from Germany

33

Total for encapsulation in BWR canisters

47,904

Total for encapsulation in PWR canisters

6,049

Total for encapsulation

53,953

The assemblies were selected to give a combined decay power of 1,700 W at the time of disposal.
That is, assemblies with a combination of burnup/age lying above the curves in Figure 5‑1 werecombined with assemblies with a combination burnup/­age below the curves. Based on the burnup
and age of the assemblies to be deposited and the assumed encapsulation period, it is not possible tofill all the assembly positions in all of the canisters if their summed up decay power shall not exceed
1,700 W. Consequently, there will be a number of canisters containing less than the maximum
possible number of assemblies. The simulations were carried out to minimise the number of unfilledassembly positions in the canisters. In Figure 5‑2 and Figure 5‑3 the resulting number of BWR andPWR canisters to be deposited is illustrated. The canisters are divided into groups with respect tothe average burnup of the encapsulated assemblies and the number of assemblies in the canisters.
A detailed account of the canisters to be deposited each year is given in Table C-3 and C-4 inAppendix C.

2000

Number of BWR canisters

1800
1600
1400
1200
1000
800
600
400
200
0

<38 MWd/kg HM>42 MWd/kg HM

38–42 MWd/kg HM

Number of assemblies
<10

10

11

12

Figure 5‑2. The BWR canisters to be deposited divided into groups with respect to the average burnup of
the encapsulated assemblies. The number of assemblies in the canister is indicated by different colours.
The total number of BWR canisters is 4,451.

TR-10-13

39

1000

Number of PWR canisters

900
800
700
600
500
400
300
200
100
0
<42 MWd/kg HM

42–47 MWd/kg HM>47 MWd/kg HM

Number of assemblies
<3

3

4

Figure 5‑3. The PWR canisters to be deposited divided into groups with respect to the average burnup of
the encapsulated assemblies. The number of assemblies in the canister is indicated by different colours.
The total number of PWR canisters is 1,652.

40

TR-10-13

6

Initial state – encapsulated spent nuclear fuel6.1

Introduction

Encapsulated spent fuel is the spent nuclear fuel encapsulated for deposition in the KBS-3 repository.
Gases and liquids in the cavities of the canister and fuel assemblies are considered as a part of theencapsulated spent fuel. The initial state of the encapsulated spent fuel refers to the properties of the
spent fuel and the gases and liquids in the cavities of the canister when the canister is finally sealed
and no further handling of the individual fuel assemblies is possible.
In this chapter the expected values of the fuel parameters given in Section 2.3 for the initial state
are presented. These values depend on the properties of the spent nuclear fuel to be deposited, theselection and drying of the fuel assemblies and their emplacement in the canister, and the change ofatmosphere in the canister described in Chapter 4.
The radionuclide inventory is an essential input to the long-term safety assessment. For the
initial state the radionuclide inventory is presented as the total inventory in the final repository(Section 6.2.2) and the inventory for a set of categories of canisters, referred to as type-canisters(Sections 6.2.3 and 6.2.4). The total radionuclide inventory in the repository will ultimately dependon the total energy output from the nuclear power plants. The inventory in each canister will, in
addition, depend on the number of assemblies in the canister, the criteria for selection of assemblies
for encapsulation and the time for encapsulation.
There are also fuel parameters of importance for the long-term safety assessment, e.g. total decay
power in a canister and content of gases and liquids in the canister, which in principal solely will
depend on the selection of assemblies for encapsulation and the handling of the assemblies. These
parameters are expected to be similar for all canisters.

6.2

Radionuclide inventory

6.2.1 Fuel parameters of importance for the radionuclide inventory
The burnup is the fuel parameter that has the largest impact on the radionuclide inventory of the
spent fuel. The burnup reflects the amount of nuclear fissions and neutron radiation that has occurredin the assembly and, thus, the content of fission and activation products and transuranium elements.
High burnup leads to higher content of fission products and, due to neutron capture, higher content
of activation products and altered content of transuranium elements. The radionuclide inventory is tosome extent also affected by:•

the irradiation and power history of the assemblies, i.e. how long time and to which power they
have been utilised in the reactor,•

the fuel type, i.e. UOX or MOX.

When the construction materials in the fuel assemblies are exposed to neutron radiation in the reactor
vessel, neutron capture will lead to formation of activation products. The amount and compositionof the construction materials and the length of time the fuel assembly has been in the reactor will
determine the content of activation products in the construction materials.
The power history, i.e. the power developed per length of fuel rod or fuel assembly over the irradiation period in the reactor vessel, referred to as the linear heat generation rate, is strongly correlated
to the fission gas release (FGR). The FGR in turn is used to determine the part of the radionuclide
inventory that is located at the fuel grain boundaries and in the gap between the fuel and the cladding /­Werme et al. 2004/. This part of the inventory is referred to as the gap inventory and will incomparison to the radionuclides embedded in the fuel matrix be released very rapidly if the spent
fuel pellets are exposed to vapour or water.
The type of fuel, UOX or MOX, will affect the content of transuranium elements.
The radionuclide inventory will get altered as the radioactive decay proceeds; consequently the ageof the fuel assemblies will also impact the radionuclide inventory.
TR-10-1341

6.2.2 Total radionuclide inventory in the final repository
As stated in Section 6.1, the total radionuclide inventory in the final repository will ultimately
depend on the total energy output from the nuclear power plants. As discussed in Section 2.2.1,
the energy can be provided by several assemblies with low burnup or fewer with high burnup. The
parameters determining the number of spent fuel assemblies a nuclear power plant will generate are:•the thermal power of the nuclear power plant,•

the number and type of assemblies in the reactor core,•

the target burnup of the assemblies, in which the accessibility and utilisation of the power plant is
considered,•the operation time of the nuclear power plant.

The presented total radionuclide inventory is based on the reference scenario for the operation of
the nuclear power plants presented in Section 2.1.1, which will result in the total number of BWR
and PWR assemblies presented in Table 2‑2. From the total number of assemblies and their summed
burnup, the average burnups of a BWR and PWR assembly have been calculated to 40.4 and44.8 MWd/kgU, respectively. The total radionuclide inventory was calculated for the average age ofthe fuel assemblies in 2045, i.e. the year the last nuclear power plant is closed down according to thereference scenario.
As discussed in Section 6.2.1, the burnup is the fuel parameter that has the largest impact on theradionuclide inventory of a spent fuel assembly. The content of activation products is also, to someextent, affected by the amount and composition of the construction materials and the irradiation
history. The impact of the construction materials on the content of fission products and transuranium
elements can be neglected. The content of transuranium elements is though affected by the irradiation history, but the burnup is the main parameter determining the inventory. Further, the relationshipbetween the content of transuranium elements in an assembly and its burnup is not linear, but thegeneration of transuranium elements will be higher for the last produced MWh than for the first.
Thus, calculating the total inventory from the mean burnup will result in a certain underestimationof the transuranium element inventory, also see Section 6.2.6.
The inventory of radionuclides and stable isotopes was calculated for the typical BWR and PWR
assemblies presented in Section 2.3.2. The calculated total radionuclide inventory has been divided
into the inventory in the fuel pellets, i.e. the fuel matrix, and the inventory in the construction materials, i.e. all other materials in the assembly. The inventory in the spent fuel matrix was calculated forthe average BWR and PWR burnups with the computer program Origen-S /­SKBdoc 1221579/. The
inventory of activations products was estimated from calculations with the computer program IndAct
and CrudAct /­SKBdoc 1179234/. In the estimates of the activation product inventory, the burnup was
assumed to be the same as for the fuel matrix. The resulting total inventory in the BWR and PWR
assemblies of thirteen radionuclides of importance for the radiotoxicity and calculated long-term risk
are given in Table 6‑1. The total inventory of all radionuclides is given in Table C-2 in Appendix C.
When calculating the total radionuclide inventory in the construction materials, it was assumed that
one out of four PWR assemblies included a control rod cluster (see Figure 2‑5). In addition, someassemblies will include inserts, such as start-up neutron sources, boron glass rods and plugs. Since
these assemblies are very few, these inserts were not included in the calculated inventory in theconstruction materials. Note that the fuel channels, see Figure 2‑5, have been removed from 1,520
of the BWR assemblies.
The calculated average burnup of the BWR and PWR assemblies given in Table 6‑1 are based on
the total number of assemblies from the reference scenario for the operation of the nuclear power
plants. However, the radionuclide inventory of the 83 BWR MOX assemblies used (3 assemblies),or to be used (80 assemblies), in Oskarshamn has been calculated separately assuming a burnup of
50 MWd/kgHM.
The radionuclide inventories of the miscellaneous fuels accounted for in Section 2.1.2 have not been
analysed in detail. For these fuels, assumptions concerning the radionuclide inventories that will
result in overestimations have been made according to Table 6‑2.

42

TR-10-13

Table 6‑1. The total inventory in BWR and PWR assemblies of thirteen radionuclides (Bq) of
importance for radiotoxicity and calculated long-term risk (alphabetic order). The inventories
are calculated for the calendar year 2045.
Radionuclide

Radionuclide inventory (Bq)
BWR assemblies
47,637 BWR UOX
40.4 Mwd/kgU
Fuel

1
2

PWR assemblies
1

Constr.
materials/
crud

267 BWR MOX
50 MWd/kgHM
Fuel

2

Constr.
materials/
crud6,016 PWR UOX
44.8 MWd/kgU
FuelConstr.
materials/
crud

Am-241

1.2E+18

1.1E+11

4.8E+16

2.4E+09

4.9E+17

6.6E+10C-14

1.1E+142.9E+14

6.2E+112.1E+12

3.4E+13

6.9E+1333 PWR MOX
34 MWd/kgHM
Controlrods

FuelConstr.
materials/
crud

4.5E+15

3.4E+08

1.7E+18

1.5E+139.7E+10

3.3E+115.1E+14

Total
activity

Cl-36

1.8E+122.8E+10

1.2E+102.1E+08

4.9E+118.0E+09

4.5E+09

1.4E+09

3.8E+07

2.3E+12

Cs-137

1.7E+19

5.2E+128.5E+16

2.7E+10

6.3E+18

1.1E+12

0.0E+00

1.0E+16

3.3E+09

2.3E+19

2.6E+11

1.7E+108.9E+14

1.3E+128.8E+07

4.3E+129.3E+14

1.6E+15

1.7E+08

1.3E+18

0.0E+00

2.2E+14

1.5E+09

1.4E+17I-129

1.0E+13

0.0E+00

9.6E+10Nb-94

4.6E+10

3.6E+13

6.2E+08

3.8E+121.3E+10

Pu-2389.5E+17

9.7E+102.1E+16

1.3E+09

3.7E+17

6.0E+10

Pu-2391.0E+17

9.2E+112.7E+15

5.3E+09

3.8E+16

2.8E+11

1.4E+13

Pu-2401.9E+17

1.2E+10

6.4E+15

3.2E+08

6.1E+16

8.2E+09

5.9E+14

3.9E+072.5E+17

Pu-241

7.4E+18

6.4E+11

1.3E+17

6.3E+09

3.0E+18

4.0E+119.3E+15

7.1E+08

1.1E+19
1.6E+19

Sr-90

1.1E+19

4.8E+122.8E+16

2.5E+10

4.2E+18

1.0E+12

3.6E+15

3.1E+09

U-234

4.0E+14

3.5E+073.8E+12

3.5E+05

1.9E+142.3E+07

0.0E+00

3.3E+11

1.1E+05

6.0E+14U-238

9.7E+13

7.9E+06

6.5E+124.3E+04

3.2E+13

4.4E+06

1.0E+112.5E+04

1.3E+14

Includes 222 assemblies from Ågesta.
Includes 184 Swap BWR MOX.

Table 6‑2. Assumed radionuclide inventories in the miscellaneous fuels.
Fuel type

NumberRadionuclide inventory

Spent fuel from Ågesta

222

Set to the same as in the average burnup BWR assembly.

Swap BWR MOX fuel assemblies

184

Set to the same as for the BWR MOX assemblies fromOskarshamn.

Swap PWR MOX

33

Matrix inventory calculated for a swap MOX assembly witha burnup of 34 MWd/kgHM.

Special boxes with fuel residues
from Studsvik

25

Set to the same as in the average burnup PWR assembly.

The assumptions in Table 6‑2 will result in an overestimation of the radionuclide inventories of themiscellaneous fuels since they contain a smaller amount of heavy metal and have an essentiallylower burnup than the average BWR and PWR assemblies and the BWR MOX assemblies fromOskarshamn. However, since the miscellaneous fuels comprise in total 464 assemblies out of about
54,000 assemblies, the impact of their divergent inventories on the total inventory can be neglected.

6.2.3 The type canister approach
At the time for the closure of the final repository when the encapsulation and deposition is finished,
the burnup, irradiation and power history and age of the assemblies in each canister will be known
and the radionuclide inventory can be calculated for each individual canister. However, at thisstage it is not reasonable to calculate the inventory in individual canisters. Therefore, a set of
type-canisters has been defined based on the assumption that the criteria for maximum allowed
total decay power in a canister will restrict the possible variation in radionuclide inventory. The
type‑canisters shall provide a representative and adequate description of the canisters’ content offuel, its burnup and age and the resulting radionuclide inventory in each canister.

TR-10-1343

The radionuclide inventory in each canister will depend on:•

the number of assemblies in the canister,•

the burnup of the assemblies,•

the age of the assemblies when they are encapsulated.

The burnup and the number of assemblies will be the parameters of most importance for the radio­
nuclide inventory. In a long-term perspective, the age of the fuel at deposition is of minor importancefor the radionuclide inventory since the short lived nuclides of importance for the decay power will
successively decay and no longer remain in the canister.
The part of the inventory located at the fuel grain boundaries and in the gap between the fuel andthe cladding is correlated to the fission gas release and power history. This part of the radionuclide
inventory is discussed in Section 6.3.
To illustrate the range of burnup and, thus, radionuclide inventory, a set of canisters with reasonable
combinations of burnup/age of the assemblies have been selected so that the total decay power in thecanisters will not exceed 1,700 W. The ages and burnups of the assemblies are based on the resultsfrom the simulation of the encapsulation presented in Section 5.2. The set of canisters and their
total activity are presented in Table 6‑3. The inventory in the fuel matrix of thirteen radionuclides
of importance for decay power, radiotoxicity and calculated long-term risk are given in Table 6‑4.
As can be seen from Table 6‑3 and Table 6‑4 both the total activity content and radionuclide inventory varies within the same order of magnitude. For full PWR canisters, the average and combined
canisters have similar total activity and radionuclide inventory. This illustrates that the decay power
criterion will restrict the variation in radionuclide content.
Table 6‑3. The total activity in a set of reasonable BWR and PWR canisters with a total decay
power of 1,700 W at time for encapsulation /­SKBdoc 1193244/.
Canister

Number ofassemblies

Burnup(MWd/kgU))

Age of assemblies(years)Total activity(1016 Bq/canister)

BWR low

12

30.7

20

2.1

BWR average

12

40.4

37

1.6

BWR high a

12

47.8

48

1.4

BWR high b

12

57

60

1.2

BWR unfilled

9

47.8

32

1.6

BWR-MOX

11

37.7

43

1.4

1

50

50

PWR low

4

34.2

20

2.0

PWR average

4

44.8

38

1.6

PWR high

4

57

55

1.3
1.7

PWR combination a

PWR combination b

PWR-MOX

44

1

57

20

3

34.2

40

2

57

51

1

57

20

3

44.8

32

1

34.8

57

1.5

1.6

TR-10-13

Table 6‑4. The inventory in the fuel matrix of thirteen radionuclides (Bq/canister) of importancefor decay power, radiotoxicity and calculated long-term risk (alphabetic order) in the selectedcanisters at time for encapsulation /­SKBdoc 1193244/.
Radionuclide inventory in fuel matrix for BWR canisters (Bq/canister)Radionuclide
BWR low
BWR average BWR high a
BWR high b

BWR unfilled

BWR-MOX

Am-241

1.84E+14

2.99E+14

3.49E+14

3.84E+14

2.34E+14

4.48E+14C-14

2.71E+10

2.84E+10

3.01E+10

4.79E+10

2.74E+10

2.59E+10

Cl-36

3.16E+08

4.54E+08

5.76E+08

1.67E+06

4.32E+08

4.23E+08

Cs-137

4.81E+15

4.21E+15

3.80E+15

3.38E+15

4.19E+15

3.48E+15

I-129

1.87E+09

2.54E+09

3.06E+09

3.71E+09

2.30E+09

2.51E+09

Nb-94

7.83E+06

1.16E+07

1.48E+07

1.88E+07

1.11E+07

1.20E+07

Pu-238

1.46E+14

2.38E+14

3.11E+14

3.87E+14

2.66E+14

2.59E+14

Pu-239

2.49E+13

2.52E+13

2.49E+13

2.45E+13

1.87E+13

3.30E+13

Pu-2403.62E+13

4.67E+13

5.36E+13

6.06E+13

4.01E+13

6.43E+13

Pu-241

3.28E+15

1.80E+15

1.14E+15

6.83E+14

1.92E+15

1.69E+15

Sr-90

3.51E+15

2.81E+15

2.37E+15

1.96E+15

2.68E+15

2.23E+15

U-234

9.58E+10

1.02E+11

1.15E+11

1.38E+11

7.46E+10

1.11E+11

U-238

2.46E+10

2.44E+10

2.43E+10

2.40E+10

1.82E+10

2.44E+10

Radionuclide inventory in fuel matrix for PWR canisters (Bq/canister)Radionuclide
PWR low
PWR average
PWR high
PWR com a

PWR com b

PWR-MOX

Am-241

1.96E+14

3.26E+14

3.93E+14

2.64E+14

2.64E+14

3.67E+14C-14

1.64E+10

2.26E+10

3.03E+10

1.99E+10

2.28E+10

1.99E+10

Cl-36

2.38E+08

3.23E+08

4.29E+08

2.86E+08

3.22E+08

2.85E+08

Cs-137

4.73E+15

4.10E+15

3.48E+15

4.17E+15

3.85E+15

3.86E+15

I-129

1.84E+09

2.50E+09

3.27E+09

2.20E+09

2.45E+09

2.27E+09

Nb-94

7.79E+06

1.15E+07

1.61E+07

9.87E+06

1.21E+07

1.13E+07

Pu-238

1.60E+14

2.45E+14

3.53E+14

2.18E+14

2.98E+14

2.41E+14

Pu-2392.47E+13

2.54E+13

2.53E+13

2.48E+13

1.90E+13

2.56E+13

Pu-2403.29E+13

4.08E+13

4.96E+13

3.69E+13

3.70E+13

4.86E+13

Pu-241

3.47E+15

1.87E+15

9.22E+14

2.23E+15

1.80E+15

2.19E+15

Sr-90

3.44E+15

2.72E+15

2.10E+15

2.81E+15

2.39E+15

2.49E+15

U-234

1.25E+11

1.27E+11

1.46E+11

1.27E+11

9.75E+10

1.02E+11

U-238

2.16E+10

2.14E+10

2.12E+10

2.15E+10

1.59E+10

1.92E+10

6.2.4 The type canisters and their radionuclide inventory
Based on the results presented in Table 6‑3 and Table 6‑4, demonstrating that the allowed total decay
power in a canister will restrict the possible variation in radionuclide inventory and the canisters tobe deposited according to the results of the simulation of the encapsulation presented in Section 5.2,
the following type canisters have been selected as a basis for descriptions of the radionuclide inventory in individual canisters.
•

BWR I: representing the largest part of the BWR canisters and of all deposited canisters (BWR
average).

•

BWR II: representing the high end radionuclide inventory in BWR canisters (BWR high a).

•

BWR III: representing the unfilled BWR canisters (BWR unfilled).

•

BWR-MOX: representing BWR canisters containing MOX assemblies (BWR-MOX).

•PWR I: representing the largest part of the PWR canisters (PWR average).

•PWR II: representing the high end radionuclide inventory in PWR canisters (PWR high).

•PWR III: representing the unfilled PWR canisters (PWR combination b).

•

PWR-MOX: representing PWR canisters containing MOX assemblies (PWR-MOX).

TR-10-1345

The complete calculated radionuclide inventories for the type canisters are provided in Appendix C.
Appendix C also includes the contribution to the total decay power in the canisters for the predominant radionuclides.
The selection of type canisters is made based on the burnup of the assemblies since it is the main
parameter determining the radionuclide inventory. The BWR I and PWR I canisters have been
selected since they represent the average canisters resulting from the simulation of the encapsulationof the assemblies to be deposited. The high burnup BWR II and PWR II canisters have been selected
since they represent the high end canisters with respect to radionuclide inventory. The BWR III andPWR III canisters have been selected to represent the unfilled canisters, which are the result of thecurrent decay power criterion and assumed encapsulation period. Finally, the BWR- and PWR MOX
canisters have been selected to represent the canisters containing MOX assemblies.
In the simulation of the encapsulation the resulting total number of canisters is 6,110 comprising
4,451 BWR canisters, 1,652 PWR canisters and 7 canisters with fuel residues from Studsvik, see
Section 5.2. The total numbers of the different type canisters is illustrated in Figure 6‑1 and theirradionuclide inventory is presented and discussed in the next section.
The BWR I type canister

The radionuclide inventory in the BWR I type-canister is set to the inventory in the canister
denominated “BWR average” in Table 6‑3 and Table 6‑4. The average burnup of the assemblies inthis canister is 40.4 MWd/kgU and the radionuclide inventory is regarded to be representative for all
canisters where 12 BWR assemblies with different burnup and ages have been combined so that their
total decay power conform to the decay power criterion (see Section 4.4.1) and their average burnup
lies in the interval 38–42 MWd/kgU.
The BWR I type canister also represents the radionuclide inventory in canisters with 12 assemblies
with an average burnup of the assemblies less than 38 MWd/kgU. Assuming the same radionuclide
inventory in the canisters with an average burnup of less than 38 MWd/kgU as in the BWR average
canister will result in an overestimated but still adequate inventory.
A summary of the canisters for which the BWR I canister is considered to provide an adequate
description of the radionuclide inventory is given in Table 6‑5. The radionuclide inventory in the
BWR I canister is accounted for in Appendix C.

Number of canisters

2500

2000

1500

1000

BWR
PWR

500

O
-M

PW
R

III
PW
R

II
PW
R

PW
R

I

O
M

BW
R

III
BW
R

II
BW
R

BW
R

I

X

X

0

Figure 6‑1. The total numbers of the different kinds of type canisters.

46

TR-10-13

Table 6‑5. The number of canisters for which the BWR I type canister provides an adequate
description of the radionuclide inventory.
Radionuclide inventory
Representative
Overestimatedbut adequate

Total

Number of canisters

547

1,661

2,208

Part of BWR canisters (4,451)

12%

37%

50%

Part of all canisters (6,110)

9%

27%

36%

The BWR II type canister

The radionuclide inventory in the BWR II type-canister is set to the inventory in the canister
denominated “BWR high a” in Table 6‑3 and Table 6‑4. The average burnup of the assemblies inthis canister is 47.8 MWd/kgU. With respect to the total inventory of assemblies to be deposited, theapplied criteria for selection of assemblies and the assumed period for encapsulation and deposition,
the radionuclide inventory in the BWR II canister is regarded as the high end of the BWR canisters
to be deposited.
The radionuclide inventory in the BWR II canister represents all canisters where BWR assemblies
with different burnup and age have been combined so that their total decay power is 1,700 W andtheir average burnup is at least 42 MWd/kgU. In most of these canisters, the average burnup of theassemblies will be less than 47.8 MWd/kg. The number of canisters for which the BWR II canister
is considered to provide an adequate description of the radionuclide inventory is given in Table 6‑6.
The radionuclide inventory in the BWR II canister is accounted for in Appendix C.
Table 6‑6. The number of canisters for which the BWR II type canister provides an adequate
description of the radionuclide inventory.
Representative
radionuclide inventory

Number of canisters

321

Part of BWR canisters (4,451)

7%

Part of all canisters (6,110)

5%

The BWR III type canister

The radionuclide inventory in the BWR III type-canister is set to the inventory in the canister
denominated “BWR unfilled” in Table 6‑3 and Table 6‑4. The average burnup of the assemblies in
the BWR III type canister is 47.8 MWd/kgU. The BWR III canister represents all BWR canisters
with 11 or fewer assemblies. The bulk of these canisters will have an average burnup higher than
42 MWd/kgU.
The content of short lived fission and activation products mainly depends on the burnup and age ofthe assemblies and will be similar as for the full canisters. However, the content of transuranium
elements and isotopes with long half lives will mainly depend on the encapsulated mass of uranium
and will, thus, be lower than in the canisters that contain 12 assemblies. The average number ofassemblies in the unfilled BWR canisters is 8.7 assemblies. Seen over the full set of unfilled canisters, the radionuclide inventory in the BWR III canister is adequate and slightly overestimated. The
radionuclide inventory in the canisters containing ten or eleven assemblies will be underestimatedand the inventory in canisters containing fewer than nine assemblies will be overestimated but stilladequate. The number of canisters containing eleven, ten, nine or less than nine assemblies is given
in Table 6‑7.

TR-10-13

47

Table 6‑7. The number of canisters containing eleven, ten, nine or less than nine assemblies forwhich the BWR III canisters provides an adequate description of the radionuclide inventory.
Radionuclide inventory
Underestimated but adequate
11 assemblies
10 assemblies

Representative
9 assemblies

Overestimated but adequate< 9 assemblies

Total

Number of canisters

10

173

738

734

1,655

Part of BWR canisters (4,451)

0% (0.2%)

4%

17%

16%

37%

Part of all canisters (6,110)

0% (0.2%)

3%

12%

12%

27%

The BWR-MOX type canister

The radionuclide inventory in the BWR-MOX type-canister is set to the inventory in the canister
denominated “BWR-MOX” in Table 6‑3 and Table 6‑4. It is anticipated that only one MOX assembly will be encapsulated in any given canister. The burnup of the MOX assembly is set to the maximum allowed, i.e. 50 MWd/kgHM and the burnup of the remaining assemblies is 37.7 MWd/kgU.
The radionuclide inventory in the BWR-MOX canister is regarded to be representative for all BWR
canisters containing a MOX assembly.
The inventory is calculated for a potential MOX assembly from Oskarshamn. The swap MOX
assemblies from Germany all have an essentially lower burnup and also a smaller mass of heavy
metal. However, since the MOX assembly is one out of twelve, the impact on the total radionuclide
content in the canister will be limited.
A summary of the canisters for which the BWR-MOX canister is considered to provide an adequate
description of the radionuclide inventory is given in Table 6‑8. The radionuclide inventory in the
BWR-MOX canister is accounted for in Appendix C.
Table 6‑8. The number of canisters for which the BWR-MOX type-canister provides an adequate
description of the radionuclide inventory.
Representative
radionuclide inventory

Number of canisters

267

Part of BWR canisters (4,451)

6%Part of all canisters (6,110)

4%The PWR I type-canister

The radionuclide inventory in the PWR I type-canister is set to the inventory in the canister denominated “PWR average burnup” in Table 6‑3 and Table 6‑4. The average burnup of the assemblies inthis canister is 44.8 MWd/kgU and the radionuclide inventory is regarded to be representative forall canisters where four PWR assemblies with different burnup and age have been combined so that
their total decay power is 1,700 W and their average burnup lies in the interval 42–47 MWd/kgU.
The PWR I type canister also represents full canisters with an average burnup of the assemblies lessthan 42 MWd/kgU. Assuming the same radionuclide inventory in these canisters as in the PWR
average canister will result in an overestimated but still adequate inventory.
A summary of the canisters for which the PWR I canister is considered to provide an adequate
description of the radionuclide inventory is given in Table 6‑9. The radionuclide inventory in the
PWR I canister is accounted for in Appendix C.

48

TR-10-13

Table 6‑9. The number of canisters for which the PWR I type canister provides an adequate
description of the radionuclide inventory.
Radionuclide inventory
Representative Overestimatedbut adequate

Total

Number of canisters

633

391

1,024

Part of PWR canisters (1,652)

38%

24%

62%

Part of all canisters (6,110)

10%

6%

17%

The PWR II type-canister

The radionuclide inventory in the PWR II type-canister is set to the inventory in the canister
denominated “PWR high burnup” in Table 6‑3 and Table 6‑4. The average burnup of the assemblies
in this canister is 57 MWd/kgU. With respect to the total inventory of assemblies to be deposited, theapplied criteria for selection of assemblies and the assumed period for encapsulation and deposition,
the radionuclide inventory in the PWR II canister is regarded as the high end of the PWR canisters tobe deposited.
The radionuclide inventory in the PWR II canister represents all canisters where PWR assemblies
with different burnup and age have been combined so that their total decay power is 1,700 W andtheir average burnup is at least 45 MWd/kgU. In most of these canisters, the average burnup of
the assemblies will be less than 57 MWd/kg/U. A summary of the canisters for which the PWR II
canister is considered to provide an adequate description of the radionuclide inventory is given in
Table 6‑10. The radionuclide inventory in the PWR II canister is accounted for in Appendix C.
Table 6‑10 The number of canisters for which the PWR II type canister provides an adequate
description of the radionuclide inventory.
Representative
radionuclide inventory

Number of canisters

38

Part of PWR canisters (1,652)

2%Part of all canisters (6,110)

1%

The PWR III type canister

The radionuclide inventory in the PWR III type-canister is set to the inventory in the canister
denominated “PWR combination b” in Table 6‑3 and Table 6‑4. The PWR III canister represents all
PWR canisters with three assemblies. Based on the results from the simulation of the encapsulation,
there are no PWR canisters with less than three assemblies. The average burnup of the assemblies inthis canister is 57 MWd/kgU. The bulk of the canisters with three assemblies will have an averageburnup lower than this.
The content of short lived fission and activation products mainly depends on the burnup and age ofthe assemblies and will be similar as for the full canisters. The content of transuranium elements andisotopes with long half lives will mainly depend on the encapsulated mass of uranium and will, thus,
be lower than in the canisters that contain four assemblies. The number of canisters containing threeassemblies is given in Table 6‑11.
Table 6‑11 The number of canisters containing three assemblies for which the PWR III typecanister provides an adequate description of the radionuclide inventory.
Representative radionuclide
inventory and total

Number of canisters

557

Part of PWR canisters (1,652)

34%

Part of all canisters (6,110)9%

TR-10-1349

The PWR-MOX canister

The radionuclide inventory in the PWR-MOX type-canister is set to the inventory in the canister
denominated “PWR-MOX” in Table 6‑3 and Table 6‑4. Each PWR-MOX canister contains one swap
PWR MOX assembly. The burnup of the MOX assembly is set to 34.8 MWd/kgHM, which is the
maximum burnup of the swap PWR MOX assemblies. The burnup of the remaining three assemblies
is 44.8 MWd/kgU, i.e. the average PWR assembly burnup. The radionuclide inventory in the PWRMOX canister is regarded to be representative for all PWR canisters containing a MOX assembly.
A summary of the canisters for which the PWR-MOX canister is considered to provide an adequate
description of the radionuclide inventory is given in Table 6‑12. The radionuclide inventory in the
BWR-MOX canister is accounted for in Appendix C.
Table 6‑12 The number of canisters for which the BWR-MOX type-canister provides an adequate
description of the radionuclide inventory.
Representative
radionuclide inventory

Number of canisters

33

Part of PWR canisters (1,652)

2%Part of all canisters (6,110)1%

6.2.5 Comparison of inventory in type canisters with total inventory
To inspect whether the type canisters provide a representative and adequate description of theradionuclide inventory in the canisters, the summed up total inventory in all type-canisters has been
compared with the total radionuclide inventory calculated for all assemblies in the reference scenario
for the operation of the nuclear power plants. The resulting total inventories of thirteen radionuclides
of importance for decay power, radiotoxicity and calculated long-term risk and the summed up
inventory of all nuclides are given in Table 6‑13. The table includes the summed up inventory in all
the different type canisters as well as in all canisters in the repository independently of type. The fullinventories are given in Appendix C.
Table 6‑13 The total inventories of thirteen radionuclides of importance for decay power,
radiotoxicity and calculated long-term risk in all type canisters and all canisters in the repository(calculated for the year of encapsulation) compared to the total inventory calculated for thecalendar year 2045 for the reference scenario for the operation of the nuclear power plants.
Radionuclide

BWR I

BWR II

BWR IIIBWRMOXPWR I

PWR IIPWR IIIPWRMOX

Total inall type
canisters

Total for thereferencescenario1

Am-241

6.6E+17

1.1E+17

3.9E+17

1.2E+17

3.3E+17

1.5E+16

1.5E+17

1.2E+16

1.8E+18

1.7E+18

C-14

2.2E+143.7E+13

1.5E+142.6E+13

7.3E+13

3.3E+12

3.6E+132.2E+125.5E+145.1E+14Cl-36

1.0E+12

1.9E+11

7.3E+11

1.1E+11

3.4E+11

1.7E+10

1.8E+11

9.6E+09

2.6E+122.3E+12

Cs-1379.3E+18

1.2E+18

6.9E+18

9.3E+17

4.2E+18

1.3E+17

2.1E+18

1.3E+17

2.5E+19

2.3E+19I-129

5.6E+129.8E+11

3.8E+12

6.7E+11

2.6E+121.2E+11

1.4E+127.5E+10

1.5E+13

1.4E+13Nb-94

2.0E+13

3.4E+121.3E+13

2.3E+12

6.1E+142.6E+132.8E+14

1.9E+139.8E+149.3E+14

Pu-2385.3E+17

1.0E+17

4.4E+17

6.9E+16

2.5E+17

1.3E+16

1.7E+178.0E+15

1.6E+18

1.3E+18

Pu-2395.6E+16

8.0E+15

3.1E+16

8.8E+15

2.6E+16

9.6E+14

1.1E+16

8.5E+14

1.4E+17

1.4E+17

Pu-2401.0E+17

1.7E+16

6.6E+16

1.7E+16

4.2E+16

1.9E+15

2.1E+16

1.6E+15

2.7E+17

2.5E+17

Pu-241

4.0E+18

3.7E+17

3.2E+18

4.5E+17

1.9E+18

3.5E+16

1.0E+18

7.2E+16

1.1E+19

1.1E+19

Sr-90

6.2E+18

7.6E+17

4.4E+18

6.0E+17

2.8E+18

8.0E+16

1.3E+18

8.2E+16

1.6E+19

1.6E+19U-234

2.2E+143.7E+13

1.2E+143.0E+13

1.3E+145.5E+125.4E+13

3.4E+12

6.1E+14

6.0E+14U-238

5.4E+13

7.8E+123.0E+13

6.5E+122.2E+138.0E+118.8E+12

6.3E+11

1.3E+14

1.3E+141

Detailed information in Table 6-1 and Appendix C.

50

TR-10-13

The total inventory for the reference scenario for the operation of the nuclear power plants has been
calculated for the age of the spent fuel in the year the last reactor will be taken out of operation,i.e. 2045. The encapsulated assemblies have been selected for encapsulation based on their decay
power and will have different ages. The ages will depend on the burnup and age of the assemblies
the year they were selected for encapsulation. The span of ages of the encapsulated assemblies at
encapsulation will be in the range 20–50 years. The fact that different ages have been used whencalculating the total inventory will impact the inventory of nuclides with half-lives of about 30 years
or shorter. In the long-term time perspective only radionuclides with long half-lifes will remain in
the repository. Consequently, regarding the inventory of nuclides remaining in the long-term time
perspective, the ages at encapsulation will be of minor importance. The inventories of long-lived
nuclides at encapsulation will mainly depend on the burnup and amount of heavy metal.
As indicated by the results given in Table 6‑13, it can be concluded that the total radionuclide
inventory calculated solely on the basis of the spent fuel assemblies to be deposited and the summed
radionuclide inventory in the deposited canisters as calculated based on the selected type canisters
are similar.

6.2.6

Uncertainties in the calculated fuel matrix radionuclide inventories

Origen-S

The computer program Origen (Oak Ridge isotope generation) used for calculating the radionuclide
inventory in the fuel matrix is developed as part of the code package Scale (Standardised computer
analyses for licensing evaluation). The program package which contains input data libraries andpreparation codes as well as simulation codes is regularly developed and maintained according toquality assurance procedures approved by U.S. Nuclear Regulatory Commission and Department ofEnergy /­SKBdoc 1179234/.
The uncertainties in the code can be estimated by comparison to measured activities. Comparisons
between nuclide contents calculated with Origen and measured contents are compiled in /­SKBdoc
1222975/. The results are given in Table 6‑14 the agreement is good and within the uncertainty
interval of the measured data.
Uncertainties in data

The uncertainties in the calculated inventories are mainly related to the used input data. The calculated inventories are based on four kinds of input:•

decay constants,•

fission product yield,•

when calculating dose rates conversion factors to kerma and•cross-section libraries for neutron and gamma reactions.

The decay constants, fission product yield and conversion factors to kerma are regarded as wellknown or of minor importance for the uncertainties of the calculated inventories /­Håkansson 2000/.
Table 6‑14 Comparison between nuclide contents calculated with Origen and measured contents.
Fuel type

NuclideMeaseured/calculated
mean inventories

PWR

U and Pu

1.01

Other actinides

1.11

Fission products

1.01

U and Pu

1.02

Other actinides

0.94

Fission products

0.93

BWR

TR-10-13

51

The used cross-section library will impact the results. The used libraries are regarded as representative for modern fuel types /­SKBdoc 1179234/. Further, the input data is compiled as detailed andsimilar to the actual conditions as possible. Axial differentiated enrichment, power density andburnup are considered. The power history is differentiated for the cycles in the reactor with declining
power in the assemblies and utilisation periods with intervening decay intervals to simulate therevision periods with interruption in the operation of the power plants. The same input are used in/­SKBdoc 1179234/ and /­SKBdoc 1222975/.
To estimate the uncertainties in the radionuclide inventory a parameter study was performed/­SKBdoc 1179234/. In the study the enrichment, power density and cross-section library was altered.
In addition, for PWR the content of Gd (gadolinium) in the assemblies has been altered. In /­SKBdoc
1077122/ the radionuclide inventories calculated by Origen-S are considered as realistic and wellreflecting the real inventories, also for high burnups. The content of fission products, e.g. Cs-137, isprimarily determined by the burnup and therefore the uncertainties in the calculations are consideredsmall, typically ±5%. The production of transuranium elements is more sensitive to the studied
parameters and the uncertainties are estimated to ±20%. However, the content of transuranium
elements is strongly correlated to the burnup, i.e. alterations of burnup affects the content more thanalterations of the studied parameters. In /­SKBdoc 1179234/ it was concluded that the uncertainties
can be handled by selecting conservative burnups.
Based on the discussions in /­SKBdoc 1179234/ the radionuclide inventories in the fuel matrix presented in this report can be regarded as representative for the presented scenarios for the operationof the power plants and encapsulation. The total radionuclide inventory in the final repository has
been calculated for assemblies with mean burnup. Since the number of assemblies with lower andhigher burnup respectively is similar this will yield a reasonable estimation of the total inventory.
However, the inventory of transuranium elements is to some extent underestimated. This is due tothat the relationship between the content of transuranium elements in an assembly and its burnup is
not linear, but the generation of transuranium elements will be higher for the last produced MWhthan for the first.
The inventory in individual canisters has been calculated based on the burnup of the assemblies, also
in this case the presented radionuclide inventories can be regarded as representative for the presented
scenarios. Following the recommendation in /­SKBdoc 1179234/ conservative inventories can beyielded by assuming only high burnup canisters in the repository.

6.2.7

Uncertainties in the calculated radionuclide inventories in construction
materials and crud

IndAct and CrudAct

In the computer program IndAct the material composition, radiated mass and neutron fluxes are
specified for each part of the fuel assembly. The computer program CrudAct uses the same basis
for activation as IndAct. This results in differentiated and correct calculations of the activation. The
codes have been reviewed and used for licensing purposes. For CrudAct measusrements have beenused for validation of the code. /­SKBdoc 1179234/ regards the calculated radionuclide inventories asrealistic.
Uncertainties in data

The input to IndAct and CrudAct comprises general radio physical data, material data and reactor
specific data. Both for the IndAct and CrudAct calculations realistic data from the Oskarshamn-3
and Ringhals-3 reactors have been used for BWR and PWR assemblies respectively /­SKBdoc
1179234/.
The content of activation products will depend on the material composition of the components.
The content of Co-60 is of special interest since it is the dominant radiation source for the activation
products. The induced activity in control rods will depend on the time the rods have been inserted in
the reactor core. The calculations are based on normal use of the control rods. The amount of crudis based on measurements, and is considered to be realistic. However, the amount of crud will vary52

TR-10-13

between reactors. In /­SKBdoc 1179234/ the uncertainty in the calculated induced activity in theconstruction materials and construction materials released to the crud have been estimated to a factorof two, and multiplication with two to obtain conservative values is recommended.
The content of transuraniums in the crud is continuously inspected during the operation and can be
regarded as a mean value for the Swedish reactors /­SKBdoc 1179234/. The content of transuraniums
in the crud depends on the occurrence of damages on the fuel assemblies. There are examples ofdamages that have caused an increase of the content of transuraniums in the crud of a factor of ten.
Consequently, there are a few assemblies with up to ten times more transuraniums in the crud thanthe contents stated in this report.

6.3

Fission gas release and gap inventory

6.3.1 Fission gas release in the spent fuel assemblies
As stated in Section 6.2.1 and discussed in /Werme et al. 2004/, the fraction of the radionuclides in
the fuel/cladding gap is considered to be proportional to the fission gas release (FGR). The FGR isstrongly correlated to the linear heat generation rate, which in turn will depend on the thermal powerof the nuclear reactor, the number of assemblies and configuration of the assemblies in the reactor
core and on how the fuel assemblies are utilised during operation. High fission gas release is to be
avoided because it may cause high pressure inside the fuel rod that may cause damages on the rods.
Further, in case of a severe reactor accident, the amount of gases that potentially can escape from thereactor should be kept low.
The nuclear fuel suppliers and power companies have developed fuel performance codes that, among
other fuel data, can be used to calculate the fission gas release. The codes have been developed tosupport a safe and efficient operation of the nuclear reactors. The correlations between calculated
and measured fuel data have been investigated and the codes are validated for fuel rod performanceand licensing analysis. Such codes have been used to calculate the fission gas release for a set of
BWR and PWR reactor operational cases. The BWR cases are presented in /­Oldberg 2009/ and the
PWR cases in /­Nordström 2009/. The purpose of the calculations was to investigate the fission gas
release after the planned increases in thermal powers and burnup. The results confirm the strong
correlation between the linear heat generation and the FGR. If the linear heat generation exceeds athreshold value, the FGR will increase significantly.
The calculated average fission gas releases for the operational cases in Table 6‑15 are given in
Figure 6‑2 for the BWR cases and Figure 6‑3 for the PWR cases. The average fission gas release has
been calculated per batch of fuel assemblies after each irradiation cycle in the reactor. Uncertainties
and complete data are given in /­Oldberg 2009/ for the BWR cases and in /­Nordström 2009/ for the
PWR cases.
Table 6‑15. Operational cases for which the FGR have been calculated.
Type of reactor

ReactorOperational case

Thermal power(MW)

Batch average discharege
burnup (MWd/kgU)

BWR

KKL1

Case 1

3,600

59.8

BWR

O3

Case 2

3,292

58.0

BWR

O1

Case 3

1,375

60.4

BWR

O3

Case 4

3,292

44.3

BWR

O3

Case 5

3,900

43.9

BWR

O1

Case 6

1,375

44.9

PWR

R2

Equilibrium

2,652

59.4

PWR

R2

Cycle 33

2,652

48.4

PWR

R4

Equilibrium

3,292

59.8

PWR

R4

Cycle 26

2,775

52.4

1

Kernkraftwerk Leibstadt in Germany, a reactor with high average discharge burnups.

TR-10-13

53

5
4.5

Average FGR (%)

4
3.5

KKL Case 1

3

O3 Case 2

2.5
2

O1 Case 3

1.5

O3 Case 41

O3 Case 5

0.5

O1 Case 6

0
0

20

40

60

80

Average burnup (MWd/kgU)

Figure 6‑2. Calculated average fission gas release at the end of each cycle for the BWR cases.

9

Average FGR (%)

8
7
6
5

R2 Equilibruim

4

R2 Cycle 33

3

R4 Equilibrium

2

R4 Cycle 26

1
0
0

20

40

60

80

Average burnup (MWd/kgU)

Figure 6‑3. Calculated average fission gas release at the end of each cycle for the PWR cases. The dropin burnup and FGR for R2 (Ringhals 2) Cycle 33 is explained by that only the low burnup assemblies were
loaded in the last cycle.

In /­SKBdoc 1222975/ the relations illustrated in Figure 6‑2 and Figure 6‑3 are used to extrapolate
reactor-specific relations between average burnup and FGR. In the interpolation, the numbers ofassemblies in the reactors and their thermal powers have been considered. The relations are basedon the assumption that the FGR is correlated to the linear heat generation. The interpolated relations
between burnup and FGR have then been used to estimate the FGR of the spent fuel assemblies
included in the reference scenario for the operation of the nuclear power plants.
From the average burnup of each assembly, the reactor it has been used in, and whether it was used
before or after the increase in power, the extrapolated reactor-specific relations were used to estimate
the FGR in each individual assembly. The resulting average FGR for all BWR assemblies is 1.9%
with a standard deviation of ±1.13%. The number of BWR assemblies in different FGR intervals is
illustrated in Figure 6‑4. The resulting average FGR for all PWR assemblies is 4.3% with a standard
deviation of ±3.11%. The number of PWR assemblies in different FGR intervals is illustrated in
Figure 6‑5.

54

TR-10-13

100%

9000

90%

8000

80%

7000

70%

6000

60%

5000

50%

4000

40%

3000

30%

2000

20%

1000

10%

0

Relative cumulative frequency (%)

Number of BWR assemblies

10000

0%
0

0.5

1

1.5

2

2.5

3

3.5

4

4.5

5

FGR (%)

Figure 6‑4. The number of BWR assemblies in different FGR intervals and relative cumulative frequencyof FGR in BWR assemblies.
2500

100%

Number of PWR assemblies

2000

80%
70%

1500

60%
50%

1000

40%
30%

500

20%

Relative cumulative frequency (%)

90%

10%
0

0%
0

1

2

3

4

5

6

7

8

9

10

FGR (%)

Figure 6‑5. The number of PWR assemblies in different FGR intervals and relative cumulative frequencyof FGR in PWR assemblies.

TR-10-1355

6.3.2 The fission gas release in the type canisters
The fission gas release of the assemblies in the different type canisters has been estimated based on
the average burnup of the assemblies in the canisters. However, the FGR in an assembly is strongly
correlated to the linear power, and the linear power can vary between assemblies with similar burnup
depending on in which reactor and to which power the assembly has been utilised. With respect tothis, the following approach was applied. First, the assemblies were sorted based on their burnup.
Then the FGR in the different type canisters was calculated as the average FGR of all the BWR or
PWR assemblies in an interval of ±5 MWd/kgU from the average burnup in each type canister. The
resulting estimated FGR for the assemblies in the type canisters is given in Table 6‑16.
Note that the high burnup assemblies with potentially high FGR must be combined with assemblies
with lower burnup in order for the decay power of the encapsulated assemblies to conform to thecriterion for acceptable decay power in a canister.
For the PWR-MOX canister the FGR was not estimated since the information required to estimate
reactor-specific relations between burnup and FGR was not available for the German reactors fromwhich the PWR-MOX assemblies originate. With respect to the low burnup of the MOX assembly
and an average burnup close to that of the PWR I canister, the FGR in these canisters can beassumed to be similar or less than in the PWR I type canister.

6.4

Decay power

The decay power will be calculated for all assemblies before they are selected for encapsulation.
A margin is added to the calculated decay power to ensure that the actual decay power conform tothe criteria 1,700 W. Based on comparisons between calculated and measured decay powers, theuncertainty in calculated decay power is estimated to 2%, and the current selection of assemblies
is made so that the total calculated decay power of the assemblies in a canister does not exceed
1,650 W. The decay power of each assembly, if required, can be measured in conjunction with thedelivery to the encapsulation part of the Clink facility.
At the initial state, the decay power is expected not to exceed 1,700 W in any canister. In most of thecanisters, the decay power is expected to be slightly less than 1,700 W.

6.5

Encapsulated gases and liquids

The content of gases and liquids depends on the result of the drying and gas exchange. At this stageof development, it is not possible to provide detailed information on the actual content of gases or
liquids. However, the technique for drying is well known from countries applying dry storage, andthe drying process presented in Section 4.6 is used in many of these countries. Change of atmospherein a tight container is also a well-known technique. In summary, there is no reason to believe that thewater content in the canisters will exceed the 600 g given as a premise or that the argon content will
be below the acceptable level of at least 90%.
Table 6‑16. Estimated FGR of the assemblies in the different type canisters.
BWR canisters
Type canister
FGR (%)PWR canisters
Type canister

FGR (%)

BWR I

1.3

PWR I

2.9

BWR II

2.1

PWR II

7.8

BWR III

2.1

PWR III

7.8

BWR-MOX

2.3

PWR-MOX(2.9)

56

TR-10-13

6.6

Radiation at the canister surface

The radiation dose rate at the canister surface has been calculated in /­SKBdoc 1077122/. The
highest obtained radiation dose rate was 0.18 Gy/h on limited areas of the canister. These levels
were obtained for PWR canisters containing four assemblies with a burnup of 60 MWd/kgU and anage of 30 years. The total decay power of four such assemblies will exceed the allowed total decay
power in a canister. Since the decay power as well as the radiation are related to the radioactivity of
the assemblies, it can be concluded that the radiation dose rate on the canister surface will be well
below the acceptable 1.0 Gy/h as long as the fuel assemblies selected for encapsulation conform to
the decay power criterion.

6.7

Criticality

The assemblies must not under any circumstances be encapsulated if the criticality criteria cannot be
met. Loading curves for check of criticality have been calculated for both BWR and PWR canisters.
When calculating the loading curves, all uncertainties regarding the propensity for criticality havebeen systematically investigated according to principles that are generally applied for all handlingof nuclear fuel /­SKBdoc 1193244/. Before encapsulation, each individual assembly is checkedagainst the loading curve. If an individual assembly does not conform to the acceptance criteria inthis check, an inspection of the specific set of assemblies selected for encapsulation will be made.
If it can not be shown that the selected combination of assemblies conforms to the criticality criteria,
a new selection will be made. If it is not possible to combine an individual assembly with otherassemblies to conform to the criticality criterion it can be encapsulated alone in a canister. If this is
not sufficient, the ultimate measure will be to alter the geometry, i.e. to reconstruct the assembly.

6.8

Dimensions and other parameters of interest for thesafety assessment

A number of suppliers of nuclear fuels have been and will be used by the different nuclear power
plants. The detailed design of the assemblies can vary between suppliers, see Section 2.3.2. The fuelassemblies used in the calculations of decay power, radionuclide inventory, criticality and fission
gas release are given in Appendix A which also includes detailed information about geometries andmaterials of these kinds of assemblies.

TR-10-1357

7References

SKB’s (Svensk Kärnbränslehantering AB) publications can be found at www.skb.se/publications.
References to SKB’s unpublished documents are listed separately at the end of the reference list.
Unpublished documents will be submitted upon request to document@skb.se.

Canister production report, SKB 2010. Design, production and initial state of the canister.
SKB TR-10-14, Svensk Kärnbränslehantering AB.
Design premises long-term safety, SKB 2009. Design premises for a KBS-3V repository based on
results from the safety assessment SR-Can and some subsequent analyses. SKB TR-09-22, Svensk
Kärnbränslehantering AB.
Repository production report, SKB 2010. Design and production of the KBS-3 repository.
SKB TR-10-12, Svensk Kärnbränslehantering AB.
Underground openings construction report, SKB 2010. Design, construction and initial state of
the underground openings. SKB TR-10-18, Svensk Kärnbränslehantering AB.
Håkansson R, 2000. Beräkning av nuklidinnehåll, resteffekt, aktivitet samt doshastighet för utbränt
kärnbränsle. SKB R-99-74, Svensk Kärnbränslehantering AB (in Swedish).
Nordström E, 2009. Fission gas release data for Ringhals PWRs. SKB TR-09-26,
Svensk Kärnbränslehantering AB.
Oldberg K, 2009. Distribution of fission gas release in 10×10 fuel. SKB TR-09-25,
Svensk Kärnbränslehantering AB.
SKB, 2006. Measurements of decay heat in spent nuclear fuel at the Swedish interim storage facility,
Clab. SKB R-05-62, Svensk Kärnbränslehantering AB.
SKB, 2009. Plan 2008. Costs starting in 2010 for the radioactive residual products from
nuclear power. Basis for fees and guarantees in 2010 and 2011. SKB TR-09-23, Svensk
Kärnbränslehantering AB.
SKB, 2009. Design premises for a KBS-3V repository based on results from the safety assessment
SR-Can and some subsequent analyses. SKB TR-09-22, Svensk Kärnbränslehantering AB.
Werme L, Johnson L H, Oversby V, King F, Spahiu K, Grambow B, Shoesmith D W, 2004.
Spent fuel performance under repository conditions: A model for use in SR-Can. SKB TR-04-19,
Svensk Kärnbränslehantering AB.

Unpublished documents
SKBdoc id, version

Title

Issuer, year

1077122, 2.0Strålskärmsberäkningar för kopparkapslar innehållande BWR, MOX och PWR
bränsleelement

SKB, 2009

1172138, 1.0

Kontroll av kärnämnen inom KBS-3-systemet

SKB, 2009

1179234, 1.0

SKB-Referensrapport till SAR allmän del Kapitel 6 – Källtermer

SKB, 2009

1193244, 4.0

Criticality safety calculations of disposal canisters

SKB, 2010

1221567, 2.0

Simulering av fyllning av kapslar för slutförvaring av utbränt kärnbränsle

SKB, 2010

1221579, 2.0

Aktivitetsinnehåll i kapslar för slutförvar

SKB, 2010

1222975, 2.0

Beräkning av fissionsgasfrigörelse för bränslet i slutförvaret

SKB, 2010

TR-10-1359

Appendix A
Fuel types
Different types of fuel assemblies have been used and will be used in the operation of the Swedish
nuclear power plants. Information on the number of fuel assemblies of representative fuel types
stored in the pools in Clab and in the storage pools at the reactors as well as the fuel in the reactor
cores at the end of December 2008 is given in Table A-1. There are minor variations of the different
fuel types, all these variations are not included in Table A-1. However, the table contains a set offuel types representative for the current inventory of spent fuel assembles. The BWR and PWR
fuel assemblies used in the calculations of the radionuclide inventory, criticality, decay power andfission gas release are given in Table A-2. In Table A-3 and A-4 main parameters for fuel assemblies
representative for Swedish fuel are given.
Table A-1. Types of fuel assemblies used in Swedish reactors until the end of 2008.
Fuel assembly type

No of fuel assemblies1

BWR

8,720
1,040
120
620
730
380
190
3,380
110
480
4,660
3,140
1,970
480
750
1,200

AA8×8
Exxon 8×8
KWU 8×8-2
ANF 9×9-5
KWU 9×9-5
KWU 9×9-Q
Atrium 9A, 9B
Atrium 10B (incl XM,XP)
GE11S
GE12S
Svea 64 (incl. 64-1)
Svea 100
Svea 96 (incl. 96S/L)
Svea 96 Optima
GE14
Svea 96 Optima 2
Svea 96 Optima 3

BWR MOX

AA 8×8 (MOX O1)

PWR

W15×15
Areva 15×15
KWU 15×15
W 17×17
F 17×17
AA 17×17
F 17×17 AFA3G (incl. variants)
17×17 HTP (incl. X5,M5)

1

Comment

More will comeMore will comeMore will comeMore will come

3
370
170
640
520
890
170
270
330

More will comeTo be noted the numbers are rounded.

Table A-2. Types of fuel assemblies used in calculations for this report.
Fuel assembly typeCalculations

BWR

Svea 96 Optima 2

Decay power
Radionuclide inventory
Fission gas release

Svea 96 Optima 3

Criticality

Areva 15×15 AGORA-5A

Fission gas release

17×17 Areva AGORA-7H. (AGORA-7H
is not used in Swedish PWR:sFission gas release

15×15 AFA3G
F 17×17

Criticality calculations
Decay power
Radionuclide inventory

PWR

TR-10-13

61

62

Table A-3. Main parameters for BWR fuel assemblies /­SKBdoc 1193244/.
AA
8x8

Exxon
8x8

KWU
8x8-2

ANF
9x9-5

KWU
9x9-5

KWU
9x9-Q

Atrium9A

Atrium10B

Atrium
10 XM

Atrium10
MOX

Atrium
10 XP

GE11
S

GE12
S

GE14

GNF
2

Svea
64

Svea
100

Svea
100

Svea
96

Svea
96

Svea
96

Svea
96
OptimaNo of fuel rods
No water rods
UO2 density (g/cc)
Fuel rod pitch (mm)
Fuel rod outer diameter(mm)
Fuel rod inner diameter(mm)
Cladding thickness(mm)Cladding material
Pellet diameter (mm)

63
1
10.4
16.3
12.3

63
1
10.5
16.3
12.3

62
2
10.5
16.3
12.3

76
5
10.4
14.5
10.6

76
5
10.5
14.5
11

72
9
10.5
14.5
11

72
9
10.6
14.5
11

91
9
10.6
13
10.1

91
9
10.8
13
10.3

91
9
10.8
13
10.3

91
9
10.5
13
10.3

74
17
10.6
14.4
11.2

91
9
10.7
13
10.3

92
8
10.4
13
10.3

92
8
10.5
13
10.3

63
1
10.5
15.8
12.3

100
0
10.5
12.4
9.62

100
0
10.5
12.7
9.62

96
4
10.5
12.4
9.62

96
4
10.5
12.7
9.62

96

96

10.5
12.7
9.62

10.7

10.7

10.7

9.11

9.66

9.66

9.67

8.86

9.04

9.04

9.1

9.76

8.98

8.94

9.06

10.7

8.36

8.36

8.36

8.36

8.36

0.8

0.84

0.82

0.74

0.67

0.67

0.67

0.6

0.62

0.62

0.59

0.71

0.64

0.66

0.6

0.8

0.63

0.63

0.63

0.63

0.63

Zr2
10.4

Zr2
10.3

Zr2
10.4

Zr2
9.49

Zr2
9.5

Zr2
9.5

Zr2
9.5

Zr2
8.67

Zr2
8.87

Zr2
8.87

Zr2
8.87

Zr2
9.55

Zr2
8.81

Zr2
8.76

Zr2
8.88

Zr2
10.4

Zr2
8.19

Zr2
8.19

Zr2
8.19

Zr2
8.19

Zr2
8.19

Water rod outer
diameter (mm)Water rod innerdiameter (mm)Cladding thickness(mm)Compartment outermeasures (mm)Compartment innermeasures (mm)Compartment wall
thickness (mm)Compartment material
Central cross inner
width (mm)Central cross wall
thickness (mm)Central compartment
outer measures (mm)Central compartment
outer measures (mm)Central compartment
wall thickness
Central compartment
material
Active fuel length (mm)

12.3

12.3

15

14

13.2

20.9x
20.9

37.5x
37.5

35.15x
35.15

24.9

24.9

24.9

24.9

12.3

10.7

10.7

13.4

13.4

11.6

23.4

23.4

23.4

23.4

11.1

0.8

0.84

0.8

0.3

0.78

0.72

0.73

0.73

0.76

0.76

0.76

0.76

0.6

139

139

139

139

139

139

139

139

139

139

139

137

137

137

138

140

140

140

140

140

134

134

134

134

134

134

134

134

134

134

134

134

134

134

134

137

137

137

137

2.3

2.3

2.3

2.3

2.3

2.3

2.3

2.3

2.3

2.3

2.3

1.65

1.65

1.65

1.65

1.1

1.1

1.1

Zr

Zr

Zr

Zr

Zr

Zr

Zr

Zr

Zr

Zr

Zr

Zr

Zr

Zr

Zr

Zr
29.7

Zr
29.7

0.8

0.8

Fuel typeTR-10-13

3712

3680

3712

33.4x
33.4

33.4x
33.4

33.4x
33.4

0.8

0.8

0.8

Zr

Zr

Zr
3712

3680

3680

3712

Svea
96
Optima
2
96

Svea
96
Optima
3
96

10.5
12.7
10.3/9.6
2
8.94/
8.36

10.6
12.8
9.84

10.6
12.8
9.84

8.63

8.63

8.77/8.1
9

8.48

8.48

140

140

140

140

137

137

137

137

137

1.1

1.1

1.4

1.1

1.1

1.4

Zr
29.7

Zr
29.7

Zr
29.6

Zr
29.6

Zr
29.6

Zr
29.6

Zr
29.6

0.8

0.8

0.8

0.8

0.8

0.8

0.8

TR-10-13

Table A-4. Main parameters for PWR fuel assemblies /­SKBdoc 1193244/.
Fuel typeW15x15 KWU15x15 F15x15AFA3G 15x15AGORA

W17x17 AA17x17 F17x17 S17x17HTP 17x17 HTP X5 17x17 HTP M5 17x17 HTP X5 17x17 AFA3G

No of fuel rods

204

204

204

204

264

264

264

264

264

264

264

264

Fuel rod pitch (mm)
Fuel rod outer diameter (mm)
Fuel rod inner diameter (mm)Cladding thickness (mm)Pellet diameter (mm)Cladding material

14.3
10.72
9.48
0.62
9.29
Zr4

14.3
10.75
9.3
0.725
9.11
Zr4

14.3
10.72
9.484
0.618
9.294
M5

14.3
10.77
9.505
0.6325
9.33
Zr4

12.6
9.5
8.36
0.57
8.19
Zr2

12.6
9.5
8.36
0.57
8.19
Zr4

12.6
9.5
8.36
0.57
8.19
Zr4

12.6
9.55
8.33
0.61
8.17
Zr4

12.6
9.55
8.33
0.61
8.165
Zr4

12.6
9.5
8.364
0.568
8.192
M5

12.6
9.5
8.35
0.575
8.192
M5

12.6
9.5
8.355
0.5725
8.192
Zr4

Active fuel length (mm)

3658

3658

3658

3658

3658

3658

3658

3658

3658

3658

3658

3658

UO2 density (g/cc)*
No of guide tubes
Guide tube material
Guide tube outer diameter (mm)Guide tube inner diameter (mm)Guide tube cladding thickness (mm)No of instrument tubes
Instrument tube material
Instrument tube outer diameter (mm)Instrument tube inner diameter (mm)Instrument tube cladding thickness (mm)

10.7
20
Zr4
13.87
13.01
0.43
1
Zr4
13.87
13.01
0.43

10.7
20
Zr4
13.86
13
0.43
1
Zr4
13.89
13.03
0.43

10.7
20
M5
14.1
13.05
0.525
1
M5
14.1
13.05
0.525

10.7
20
M5
14.1
13.05
0.525

10.7
24
Zr4
12.09
11.05
0.52
1
Zr4
12.24
11.428
0.406

10.7
24
Zr4
12.24
11.44
0.4
1
Zr4
12.24
11.428
0.406

10.7
24
Zr4
12.05
11.25
0.4
1
Zr4
12.05
11.25
0.4

10.7
24
PCAm
12.24
11.3
0.47
1
PCAm
12.24
11.3
0.47

10.7
24
PCAm
12.45
11.45
0.5
1
PCAm
12.24
11.3
0.47

10.7
24
PCAm
12.45
11.45
0.5
1
PCAm
12.45
11.45
0.5

10.7
24
PCAm
12.24
11.3
0.47
1
PCAm
12.24
11.3
0.47

10.7
24
Zr4
12.45
11.45
0.5
1
Zr4
12.45
11.45
0.5

M5
14.1
13.05
0.525

* The UO2-density 10.7g/cc is used for all fuel types. This density is higher than the fabricated values.

63Appendix B
Material specification for representative fuel types for BWR
and PWR
Table B-1. Fuel matrix compostion of unirradiated fuel in one fuel assembly /­SKBdoc 1221579/.
Material data

PWR
Areva 17×17

BWR
Svea 96 Optima 2

Enrichment

4.0 (% U-235)

3.6 (% U-235)

Weight in 1 fuel assembly (kg)
Fuel composition

U-tot

464

175

U-23410.19

0.05

U-2351

18.6

6.3

U-2361

0.1

0.04

U-238

1

O

445.2

168.6

62

23

1
In the fuel matrix a content of 0.03% U234 + 0.02% U236 for BWR and 0.04% U234 + 0.02% U236 for PWR isassumed.

Material data

Atrium 10B
O3 MOX

Enrichment

4.6 (% Pufiss) + 0.2 (% U-235)

Weight in 1 fuel assembly (kg)
Fuel composition

HM-tot

174

U-235

0.3

U-238

162.3

Pu-2380.1

Pu-2397.2

Pu-240

2.9

Pu-241

0.7

Pu-2420.5

O

23

TR-10-13

65

Table B-2. Impurities in the fuel matrix.
Element

Assumed in calculations(ppm)

Representative values for fuel matrices1
(ppm)

Ag

0.05

<0.05

Al

6

3–6

B

0.05

<0.05

Bi

0.5

<0.5

Ca

3

<3

1/3 above LRV

Cd

0.233

average 0.233

min 0.2 max 0.6

Co

0.5

<0.5

Cr

1

<1

10% above LRV

Cu

0.5

average 0.5

min 0.2 max 7

F

2

<2

20% above LRV

Fe

5

<5

20% above LRV

In

0.3

<0.3

Li

0.05

<0.05

Mg

1

<1

Mn

2

<2

Mo

5

<5

N2

14

–

Ni

–

<1

Pb

0.6

<0.6

Si

10

<10

Sn

0.8

0.6–0.8

Ti

10

<10

V

0.3

<0.3

Zn

25

<25

Dy

10

<10

Eu

0.02

<0.02

Gd

0.06

<0.06

Sm

0.04

<0.04

C

8.4

average 8.4

Cl

2

2

Ni

5

5

W

0.2

0.2

20% above LRV

min 3 max 28(LRV_Lowest reported value)1
2

Personal communication Westinghouse.
Assumed in accordance with /­SKBdoc 1179234/.

66

TR-10-13

Table B-3. Components and material composition for components in a BWR fuel assembly with
fuel channel /­SKBdoc 1179234/.
Component

Fuel assembly
Cladding Spacer

Springs,
part lengthrod

Springs,
ordinaryrod

Top plateBottomwith handle plate

Fuel channel
ChannelChannelbottom piece

Material

Zry-2

X-750

AISI 302AISI 302AISI 304AISI 304L

Zry-2

AISI 304L

Weight (kg)

49.3

0.86

0.088

1.26

2.6

0.96

31.5

8.4

Material composition (%)

C

0.015

0.03

0.07

0.07

0.04

0.02

0.015

0.02

N

0.004

0.01

0.04

0.04

0.04

0.04

0.004

0.04

O

0.14

0.01

0.01

0.01

0.01

0.01

0.14

0.01

Al

0.005

0.7

0.002

0.002

0.002

0.002

0.005

0.002

Si

0.01

0.3

0.6

0.6

0.6

0.6

0.01

0.6

P

0.005

0.02

0.02

0.02

0.02

0.02

S

0.005

0.015

0.015

0.015

0.015

0.015

Cl
Ti

0.00010.00010.00010.00010.0001

0.004

2.5

0.01

0.01

0.01

0.01

0.01

0.001

0.001

0.001

0.001

0.1

16

18.5

18.5

18.5

18.5

V
Cr

0.0001
0.004

0.01

0.1

18.5

0.001

Mn

0.003

0.2

1.3

1.3

1.3

1.3

0.003

1.3

Fe

0.15

7.00

69.05

69.05

69.08

69.10

0.15

69.1

Co

0.0001

0.01

0.03

0.03

0.03

0.03

0.0001

0.03

Ni

0.05

71.99

10.00

10.00

10.00

10.00

0.05

10

Cu

0.003

0.003

As
Nb
Mo

0.0005

Sn

1.5

Sb
Ta

0.1

0.1

0.1

0.1

0.1

0.05

0.01

0.01

0.01

0.01

0.9

0.01

0.01

0.01

0.01

0.05

0.2

0.2

0.2

0.2

0.0005
1.5

0.01

0.01

0.01

0.01

0.01

0.005

0.001

0.001

0.001

0.001

0.1

0.01

0.01

0.01

0.01

0.01

0.01

0.01

0.01

0.01

0.01

Hf

0.01
98.0

0.01
98.0

0.000020.000020.000150.00015

1

0.01
0.01

0.005

U

0.2
0.001

W
Zr1
Th

0.1
0.01

0.005

0.01

The mass given for Zr is the total mass in the fuel assembly, the water cross is included.

TR-10-13

67

Table B-4. Components and material composition for components in a PWR fuel assembly/­SKBdoc 1179234/.
Component

Top nozzle
SS

Top nozzle
Zr

BottomnozzleSS

SpacerZr

SpacerinconelGuidethimbleZr

Guide
thimbleSS

Cladding
Zr

Cladding1
SS

Material

304L_1

Inc718_1

304L_2

Zry4_3

Inc718_3

Zry4_4

316L_4

M5_5

302_5

Weight (kg)

6.5

1.1

4.8

7.2

0.75

14.2

0.7

108.1

3

Material Composition (%)

Li

0.00001

C

0.02

0.03

0.02

0.015

0.03

0.015

0.025

0.015

0.12

N

0.04

0.01

0.04

0.004

0.01

0.004

0.04

0.004

0.04

O

0.01

0.01

0.01

0.14

0.01

0.14

0.01

0.14

0.01

Na

0.001

Al

0.002

0.5

0.002

0.005

0.5

0.005

0.002

0.005

0.002

Si

0.6

0.3

0.6

0.01

0.3

0.01

0.6

0.01

0.6

P

0.02

0.005

0.02

0.005

0.02

0.02

S

0.015

0.005

0.015

0.005

0.015

0.015

Cl

0.00010.00010.00010.00010.00010.0001

Ca

0.002

Ti

0.01

0.9

0.01

V

0.001

0.01

0.001

Cr

18.5

19

18.5

0.1

19

Mn

1.3

0.3

1.3

0.003

Fe

69.05

21.05

69.05

Co

0.03

0.05

Ni

10

Cu

0.1

Zn

0.01

As

0.01

Se

0.004

Nb

0.03

4.59

0.03

0.01

4.59

0.01

0.03

Mo

0.2

0.05

0.2

0.00050.05

0.0005

2.5

0.2

Ag

0.00010.00010.0001

Sn

0.01

0.01

0.01

0.01

0.01

Sb

0.001

0.005

0.001

0.001

0.001

Ce

0.01

0.01

0.01

Ta

0.01

0.51

0.01

W

0.01

0.01

0.01

Hf

0.00001

0.00001

0.001

0.00001

0.001

0.002

0.001

0.002
0.004

0.9

0.004

0.01

0.01

0.002
0.004

0.01

0.001

0.001

0.1

17

18.5

0.3

0.003

1.3

1.3

0.22

21.05

0.22

66.25

0.03

0.0001

0.05

0.0001

0.03

52.50

10

0.004

52.5

0.004

12

9

0.1

0.1

0.003

0.1

0.003

0.1

0.1

0.01

0.01

0.01

0.01

0.01
0.05

0.01

0.05

0.004

69.95
0.00001

0.004

0.0001
1.5

0.01

1.5

0.005

0.01
0.51
0.005

0.01

0.004
1

0.03

0.01
0.005

0.01

0.01
0.005

0.01

0.006

0.006

0.01

97.97

97.97

98.81

Th

0.000020.000020.00002U

0.000150.000150.00015

Zr

1

0.001

0.03

The stainless steel in the springs in the fission gas plenum.

68

TR-10-13

Table B-5. Material composition for control rod clusters in a PWR fuel assembly/­SKBdoc 1179234/.
Component

Absorber pins

Absorber pins

Top piece

Material

304_1

AgInCd_1

304_2

Weight (kg)

12.5

51.4

3.8

Material composition (%)

Li

0.00001

0.00001

C

0.07

0.07

N

0.04

0.04

O

0.01

0.01

Na

0.001

0.001

Al

0.002

0.002

Si

0.6

0.6

P

0.02

0.02

S

0.015

0.015

Cl

0.00010.0001

Ca

0.002

0.002

Ti

0.01

0.01

V

0.001

0.001

Cr

18.5

18.5

Mn

1.3

1.3

Fe

68.9

68.9

Co

0.1

0.1

Ni

10

10

Cu

0.1

0.1

Zn

0.01

0.01

As

0.01

0.01

Se

0.004

0.004

Nb

0.03

0.03

Mo

0.2

Ag

0.0001Cd

0.2
80

0.0001

5

In

15

Sn

0.01

0.01

Sb

0.001

0.001

Ce

0.01

0.01

Ta

0.01

0.01

W

0.01

0.01

Zr

0.001

0.001

TR-10-1369

Appendix C
Radionuclide inventory
List of tables
Table C-1 Radionuclide information – halflives used in Origen-S calculations and calculated specificactivities given as Bq/g and Bq/mol.
Table C-2 Total radionuclide inventory based on the number of fuel assemblies in the SKB reference
scenario. The inventories are calculated for year 2045.
Table C-3 Simulation of BWR canister encapsulation rate during year 2023–2070.
Table C-4 Simulation of PWR canister encapsulation rate during year 2023–2070.
Table C-5 Total radionuclide inventory estimated from the inventory in the type canisters at encapsulation (different times for the different type canisters).
Table C-6 Radionuclide inventory in BWR I type canister (12 assemblies, burnup 40.4 MWd/kgHM,
age 37 years).
Table C-7 Radionuclide inventory in PWR I type canister (4 assemblies, burnup 44.8 MWd/kgHM,
age 38 years).
Table C-8 Radionuclide inventory in BWR II type canister (12 assemblies, burnup 47.8 MWd/kgHM,
age 48 years).
Table C-9 Radionuclide inventory in BWR III type canister (9 assemblies, burnup 47.8 MWd/kgHM,
age 32 years).
Table C-10 Radionuclide inventory in BWR MOX type canister (11 BWR assemblies, burnup
37.7 MWd/kgHM, age 43 years and 1 MOX assembly, burnup 50 MWd/kgHM, age 50 years).
Table C-11 Radionuclide inventory in PWR II type canister (4 assemblies, burnup 57 MWd/kgHM,
age 55 years).
Table C-12 Radionuclide inventory in PWR III type canister (2 assemblies, burnup 57 MWd/kgHM,
age 51 years and 1 assembly burnup 57 MWd/kgHM, age 20 years).
Table C-13 Radionuclide inventory in PWR MOX type canister (3 PWR assemblies, burnup
44.8 MWd/kgHM, age 32 years and 1 MOX assembly burnup 34.8 MWd/kgHM, age 57 years).
Table C-14 Contribution from predominant radionuclides to the total decay power at time for
encapsulation.
Table C-15 Contribution from different isotopes in the fuel matrix to the total mass in a BWR I typecanister and PWR I type canister (for isotope masses greater than 1 g/canister).
The different ages of the fuels in the type canisters depend on the different decay times required toreach an acceptable total decay power in the canisters for different burnups.

TR-10-13

71

Table C-1. Radionuclide information – halflives used in Origen-S calculations /­SKBdoc 1221579/
and calculated specific activities given as Bq/g and Bq/mol.
NuclideHalf-life(years)Specific
activity(Bq/g)

Specific
activity(Bq/mol)

NuclideHalf-life(years)Specific
activity(Bq/g)

Specific
activity(Bq/mol)

Ac-22721.8

2.67E+12

6.07E+14

Pa-231

3.28E+04

1.75E+09

4.04E+11

Ag-108 m

127.0

9.650E+11

1.042E+14

Pa-23327.0 d

7.68E+14

1.79E+17

Am-241432.7

1.269E+11

3.059E+13

Pa-234 m

1.2 m

2.5E+19

5.8E+21

Am-24216.0 h

2.99E+16

7.25E+18

Pb-21022.3

2.83E+12

5.94E+14

Am-242m

141.0

3.879E+11

9.387E+13

Pd-107

6.50E+06

1.90E+07

2.04E+09

Am-243

7.37E+03

7.39E+09

1.80E+12

Pu-23887.7

6.34E+11

1.51E+14

C-14

5.73E+03

1.65E+11

2.31E+12

Pu-239

2.41E+04

2.30E+09

5.49E+11

Cd-113m

14.1

8.31E+12

9.39E+14

Pu-2406.56E+03

8.41E+09

2.02E+12

Cl-36

3.01E+05

1.22E+09

4.40E+10

Pu-241

14.3

3.84E+12

9.26E+14

Cm-242

162.9 d1.23E+14

2.97E+16

Pu-242

3.73E+05

1.47E+08

3.55E+10

Cm-24328.5

1.91E+12

4.64E+14

Ra-226

1.60E+03

3.66E+10

8.27E+12

Cm-24418.1

3.00E+12

7.31E+14

Se-79

2.95E+05

5.68E+08

4.49E+10

Cm-245

8.50E+03

6.36E+09

1.56E+12

Sm-151

90.0

9.74E+11

1.47E+14

Cm-246

4.73E+03

1.14E+10

2.80E+12

Sn-121m

55.0

1.99E+12

2.41E+14

Cs-1352.30E+06

4.26E+07

5.75E+09

Sn-126

1.00E+05

1.05E+09

1.32E+11

Cs-13730.0

3.22E+12

4.41E+14

Sr-90

28.1

5.23E+12

4.71E+14

Eu-152

13.3

6.55E+12

9.95E+14

Tc-99

2.11E+05

6.34E+08

6.27E+10

H-3

12.3

3.59E+14

1.08E+15

Th-2297.88E+03

7.33E+09

1.68E+12

Ho-166m

1.20E+03

6.64E+10

1.10E+13

Th-2307.54E+04

7.63E+08

1.76E+11

I-129

1.57E+07

6.54E+06

8.43E+08

Th-232

1.40E+10

4.08E+03

9.45E+05

Mo-93

3.50E+03

4.07E+10

3.78E+12

Th-23424.1 d

8.57E+14

2.00E+17

Nb-93m16.1

8.84E+12

8.22E+14

U-233

1.59E+05

3.57E+08

8.32E+10

Nb-94

2.03E+04

6.94E+09

6.52E+11

U-234

2.46E+05

2.30E+08

5.38E+10

Ni-59

7.50E+04

2.99E+09

1.76E+11

U-235

7.04E+08

8.00E+04

1.88E+07

Ni-63

100.1

2.099E+12

1.322E+14

U-236

2.34E+07

2.40E+06

5.66E+08

Np-237

2.14E+06

2.61E+07

6.19E+09

U-237

6.8 d

3.0E+15

7.1E+17

Np-23850.8 h

9.59E+15

2.28E+18

U-238

4.47E+09

1.24E+04

2.96E+06

Np-23956.5 h

8.59E+15

2.05E+18

Z-r93

1.53E+06

9.30E+07

8.65E+09m) minutes, h) hours, d) days.

72

TR-10-13

Table C-2. Total radionuclide inventory based on the number of fuel assemblies in the SKB
reference scenario. Inventory for fuel matrix based on information in /­SKBdoc 1221579/ andconstruction material, Crud and PWR control rod estimated from data given in /­SKBdoc 1179234/.
The inventories are calculated for year 2045.
Radionuclide

Ac-227
Ag-108m
Am-241
Am-242
Am-242m
Am-243
C-14
Cd-113m
Cl-36
Cm-242
Cm-243
Cm-244
Cm-245
Cm-246
Cs-135
Cs-137
Eu-152
H-3
Ho-166m
I-129
Mo-93
Nb-93m
Nb-94
Ni-59
Ni-63
Np-237
Np-238
Np-239Pa-231
Pa-233
Pa-234m
Pb-210
Pd-107
Pu-238
Pu-239Pu-240Pu-241
Pu-242
Ra-226
Se-79
Sm-151
Sn-121m
Sn-126
Sr-90
Tc-99
Th-229Th-230Th-232
Th-234
U-233
U-234
U-235
U-236
U-237
U-238
Zr-93
1
2

Radionuclide inventory (Bq)47,637 BWR1 assemblies

6,016 PWR assemblies

267 BWRMOX2 assemblies 33 PWRMOX assemblies

40.4 Mwd/kgU

44.8 MWd/kgU

50 MWd/kgHM

34 MWd/kgHM

FuelConstr.
materials
/crud

Fuel

Constr.
materials
/crud

Fuel

Constr.
materials
/crud

Fuel

Constr.
materials/crudTotal activity

6.24E+09
1.99E+12
1.18E+18
2.97E+15
2.98E+15
9.50E+15
1.13E+14
2.89E+13
1.80E+122.45E+15
3.41E+15
3.45E+17
1.27E+14
2.57E+13
1.77E+14
1.70E+19
3.90E+14
2.83E+16
3.24E+13
1.01E+13
2.58E+11
5.35E+14
4.62E+10
9.77E+11
1.03E+14
1.25E+14
1.34E+13
9.50E+15
1.05E+10
1.25E+14
9.69E+13
3.13E+08
4.27E+13
9.50E+17
1.00E+17
1.85E+17
7.39E+18
7.88E+14
1.03E+09
2.70E+13
9.85E+16
3.06E+15
1.97E+14
1.13E+19
4.98E+15
1.08E+08
1.26E+11
1.89E+05
9.69E+13
2.47E+10
4.03E+14
4.79E+12
9.88E+13
1.77E+14
9.69E+13
6.61E+14

5.09E+02
1.55E+12
1.06E+11
5.04E+08
5.07E+08
1.15E+09
2.87E+14
0.00E+00
2.82E+10
4.17E+08
2.98E+08
4.96E+10
2.14E+07
5.53E+06
0.00E+00
5.17E+12
0.00E+00
0.00E+00
0.00E+00
0.00E+00
1.18E+12
9.46E+15
3.57E+13
1.44E+15
1.53E+17
1.10E+07
2.28E+06
1.15E+09
8.44E+02
1.10E+07
7.93E+06
2.77E+01
0.00E+00
9.68E+10
9.22E+11
1.23E+10
6.39E+11
7.77E+07
8.90E+01
0.00E+00
0.00E+00
5.95E+15
0.00E+00
4.85E+12
2.03E+11
1.22E+01
1.08E+04
1.48E–02
7.93E+06
1.00E+10
3.45E+07
3.63E+05
7.47E+06
1.53E+07
7.93E+06
1.02E+14

2.43E+09
4.82E+11
4.86E+17
9.12E+14
9.16E+14
4.01E+15
3.39E+13
1.14E+13
4.86E+11
7.54E+14
1.27E+15
1.45E+17
5.79E+13
1.09E+13
5.57E+13
6.32E+18
1.10E+14
1.01E+16
1.26E+13
3.76E+12
1.24E+11
1.98E+14
1.73E+10
2.62E+11
2.68E+13
5.16E+13
4.12E+12
4.01E+15
4.07E+09
5.16E+13
3.22E+13
1.52E+08
1.61E+13
3.72E+17
3.82E+16
6.14E+16
2.97E+18
3.15E+14
4.98E+08
1.00E+13
3.98E+16
1.14E+15
7.36E+13
4.20E+18
1.84E+15
4.09E+07
6.10E+10
7.01E+04
3.22E+13
1.09E+10
1.90E+14
1.82E+12
3.67E+13
7.11E+13
3.22E+13
2.45E+14

2.71E+02
4.17E+11
6.59E+10
2.13E+08
2.14E+08
6.52E+08
6.90E+13
0.00E+00
8.03E+09
1.76E+08
1.75E+08
3.00E+10
1.52E+07
3.95E+06

4.49E+06
2.08E+09
4.77E+16
4.31E+14
4.33E+14
4.54E+14
1.93E+12
9.29E+11
3.14E+10
3.57E+14
1.89E+14
1.84E+16
3.30E+13
4.54E+122.90E+12
8.49E+16
4.76E+12
9.72E+13
5.49E+11
9.63E+10
4.19E+09
2.82E+12
1.65E+10
1.60E+12
6.17E+08
8.30E+11
1.95E+12
4.54E+14
7.06E+06
8.30E+11
5.15E+11
2.33E+06
6.42E+11
2.06E+16
2.67E+15
6.35E+15
1.34E+17
2.68E+13
7.53E+06
1.56E+11
1.87E+15
2.72E+13
2.04E+12
2.77E+16
3.21E+13
6.63E+05
9.67E+08
1.00E+02
5.15E+11
1.15E+08
3.76E+12
3.70E+09
4.33E+10
3.21E+12
5.15E+11
3.15E+12

3.60E+00
1.25E+10
2.37E+09
2.00E+07
2.01E+07
3.43E+07
2.11E+12
0.00E+00
2.07E+08
1.66E+07
7.97E+06
1.47E+09
2.12E+06
6.62E+05

1.49E+06
2.00E+09
4.52E+15
1.53E+13
1.54E+13
5.08E+13
9.66E+10
2.79E+10
1.40E+09
1.27E+13
8.70E+12
1.09E+15
1.28E+12
2.19E+11
2.29E+11
9.99E+15
2.71E+11
8.94E+12
3.84E+10
1.29E+10
5.21E+08
4.16E+11
8.85E+07
8.07E+08
6.95E+10
1.03E+11
6.94E+10
5.08E+13
2.66E+06
1.03E+11
1.04E+11
2.85E+05
8.39E+10
1.57E+15
2.18E+14
5.94E+14
9.32E+15
3.43E+12
8.48E+05
2.23E+10
1.40E+14
3.45E+12
2.73E+11
3.59E+15
4.64E+12
4.67E+04
9.67E+07
4.71E+01
1.04E+11
1.71E+07
3.32E+11
2.06E+09
1.68E+10
2.23E+11
1.04E+11
4.52E+11

1.83E+00
2.07E+09
3.36E+08
7.59E+05
7.62E+05
2.15E+06
3.25E+11
0.00E+00
3.80E+07
6.28E+05
3.78E+05
3.71E+07
3.53E+04
8.26E+03

8.67E+09
6.86E+16
1.72E+18
4.33E+15
4.34E+15
1.40E+16
5.10E+14
4.57E+15
2.34E+12
3.58E+15
4.88E+15
5.09E+17
2.19E+14
4.14E+13
2.36E+14
2.34E+19
5.05E+14
3.85E+16
4.53E+13
1.39E+13
3.24E+12
2.50E+17
9.32E+14
1.73E+15
1.85E+17
1.78E+14
1.96E+13
1.40E+16
1.46E+10
1.78E+14
1.30E+14
4.67E+08
5.95E+13
1.34E+18
1.41E+17
2.54E+17
1.05E+19
1.13E+15
1.53E+09
3.73E+13
1.40E+17
1.04E+16
2.73E+14
1.56E+19
6.86E+15
1.50E+08
1.88E+11
2.59E+05
1.30E+14
4.84E+10
5.97E+14
6.61E+12
1.35E+14
2.51E+14
1.30E+14
1.04E+15

Controlrods

6.86E+16

3.75E+12
4.53E+15
1.12E+09

1.11E+12
0.00E+00
2.15E+13
0.00E+00

0.00E+00
0.00E+00
1.56E+12
0.00E+00

1.48E+12
2.39E+17
8.92E+14
2.58E+14
2.76E+16
7.25E+06
9.63E+05
6.52E+08
4.66E+02
7.25E+06
4.42E+06
1.72E+01

1.83E+11
5.32E+13
3.21E+11
2.45E+13
2.16E+15

6.01E+10
2.79E+11
8.23E+09
3.96E+11
4.41E+07
5.64E+01
4.88E+09
0.00E+00
1.97E+14
1.04E+12
2.21E+11
6.50E+00
6.97E+03
9.41E–03
4.42E+06
2.55E+09
2.26E+07
2.45E+05
4.83E+06
9.47E+06
4.42E+06
2.62E+13

0.00E+00

0.00E+00

1.76E+09
0.00E+00
4.87E+10
0.00E+00
2.90E+10

0.00E+00

2.50E+06

2.72E+10
0.00E+00
0.00E+00
0.00E+00
8.70E+09
3.79E+13
2.58E+11
1.01E+13
1.02E+15
4.76E+04
9.06E+04
3.43E+07
4.49E+00
4.76E+04
4.35E+04
4.28E–01
1.33E+09
5.28E+09
3.23E+08
6.32E+09
2.23E+06
1.18E+00
0.00E+00
0.00E+00
3.52E+13
2.53E+10
1.47E+09
4.87E–02
1.20E+02
9.54E–06
4.35E+04
5.80E+07
3.53E+05
2.05E+02
3.75E+03
1.51E+05
4.35E+04
7.49E+11

3.30E+09
0.00E+00
3.33E+10
0.00E+00
7.03E+09
4.88E+14
4.29E+12
1.25E+12
1.14E+14
3.10E+04
3.43E+03
2.15E+06
2.79E+00
3.10E+04
2.46E+04
2.62E–01
1.70E+08
1.52E+09
3.92E+07
7.08E+08
1.76E+05
6.50E–01
2.31E+07
0.00E+00
7.14E+11
3.06E+09
1.08E+09
3.40E–02
5.37E+01
6.20E–05
2.46E+04
1.38E+07
1.14E+05
1.40E+03
2.13E+04
1.69E+04
2.46E+04
1.24E+11

Includes 222 assemblies from Ågesta.
Includes 184 Swap BWR MOX.

TR-10-13

73

BWR:
•

UO2-values from SKBdoc 1221579 (‘Nuklidlista 100320.xls’)•

Inventory for construction material from SKBdoc 1179234 by linear interpolation between 38 and
60 MWd/kg U to find inventory at 40.4 MWd/kg U.
(Data from the files ‘88_Ind-B38-000.xls’ and ‘89_Ind-B60-000.xls’)•

Inventory for Crud from SKBdoc 1179234 by linear interpolation between 38 and 60 MWd/kg U
to find inventory at 40.4 MWd/kg U.
(Data from the files ‘95_Crud-B38-000.xls’ and ‘96_Crud-B60-000.xls’)•

Average age for BWR assemblies 36.3 years and BWRmox assemblies 50 years.

PWR:•

UO2-values from SKBdoc 1221579 (‘Nuklidlista 100320.xls’)•

Inventory for construction material from SKBdoc 1179234 by linear interpolation between 30 and
60 MWd/kg U to find inventory at 44.8 MWd/kg U.
Assumption: At any given burnup or age the construction material data is the same for PWR andPWRMOX.
(Data from the files ‘90_Ind-P30-000.xls’ and ‘91_Ind-P60-000.xls’)•

Inventory for Crud from SKBdoc 1179234 by linear interpolation between 30 and 60 MWd/kg U
to find inventory at 44.8 MWd/kg U.
Assumption: At any given burnup or age the crud data is the same for PWR and PWRMOX.
(Data from the files ‘97_Crud-P30-000.xls’ and ‘98_Crud-P60-000.xls’)•

Inventory for control rods from SKBdoc 1179234 (‘110_PWR-ss.xls’). The number of control
rods assumed to be one fourth of the PWR assemblies.

•

Average age for PWR assemblies 36.9 years and PWRmox assemblies 57 years.

74

TR-10-13

Table C-3. Simulation of BWR canister encapsulation rate during year 2023–2070 /­SKBdoc
1221567/.
Year

Canisters/year
BWR

PWR

BWR
Tot

No of fuel assemblies<38 MWd
12

2023
2024
2025
2026
2027
2028
2029
2030
2031
2032
2033
2034
2035
2036
2037
2038
2039
2040
2041
2042
2043
2044
2045
2046
2047
2048
2049
2050
2051
2052
2053
2054
2055
2056
2057
2058
2059
2060
2061
2062
2063
2064
2065
2066
2067
2068
2069
2070

17
58
87
87
87
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
73
73
73
73
73
73
73
73
73
73
73
73
73
73
73
77

6
22
33
33
33
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
27
27
27
27
27
27
27
27
27
27
27
27
27
27
27
13

Total

4,451

1,652 6,103 1,661 0

1

23
80
120
120
120
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
100
100
100
100
100
100
100
100
100
100
100
100
100
100
100
90

11

38–42 MWd
10

9

<9

17
58
87
87
87
109
109
109
109
109
109
109
109
103
92
  8
  1

12

11

>42 MWd
10

9

<9

  6
17
101
108
109
109
109
109
91
25
16
13
  1

40

12

11

10

18
44
93
96
70

10

28
78

12
38
38
25
25
40

19
17
1

1

1
13
14

17

9

<9

19
52
52
70
51
3
12
41
25
11
16

43

24

54

0

0

8141

0

13

52

0

321

10

106

26
39
31
6
73
72
12
15
60

1
61
58
17

686

734

Including 267 BWR MOX.

BWR I + BWR MOX

TR-10-132475

BWR II321

BWR III

58
67
4
32
48
62
73
57
73
47
34
42

1655

75

Table C-4. Simulation of PWR canister encapsulation rate during year 2023–2070 /­SKBdoc
1221567/.
Year

Canisters/year
BWR

PWR

PWR
Tot

No of fuel assemblies

4

2023
2024
2025
2026
2027
2028
2029
2030
2031
2032
2033
2034
2035
2036
2037
2038
2039
2040
2041
2042
2043
2044
2045
2046
2047
2048
2049
2050
2051
2052
2053
2054
2055
2056
2057
2058
2059
2060
2061
2062
2063
2064
2065
2066
2067
2068
2069

17
58
87
87
87
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
109
73
73
73
73
73
73
73
73
73
73
73
73
73
73
73

6
22
33
33
33
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
41
27
27
27
27
27
27
27
27
27
27
27
27
27
27
27

23
80
120
120
120
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
150
100
100
100
100
100
100
100
100
100
100
100
100
100
100
100

2070

77

13

90

Total

4,451 1,652 6,103

1<42 MWd3

42–47 MWd
<3

6
22
33
33
33
41
39
37
32
25
22
10
4
1
1
2

4

3

>47 MWd
<3

2
4
9
16
19
31
37
40
40
39
41
41
41
41
39
40
31
31
37
31
31

4

2
1
10
10
4
10
1

32
18

23
2

12
0

0

PWR II

38

<3

9
41
9
41
39
27
27
27
27
27
27
27
27
27
27
27
27
27
26
15

1
12
391

3

1

666

1

25

0

38

532

0

Including 33 PWRMOX.

PWR I + PWRMOX

76

1057

PWR III

557

TR-10-13

Table C-5. Total radionuclide inventory estimated from the inventory in the type canisters at
encapsulation (different times for the different type canisters).
Radionuclide inventory (Bq)
Type

BWR I

BWR II

BWR III

BWR-MOX PWR I

PWR IIPWR IIIPWR-MOX TOTAL

Number ofcanisters

2,208

321

1,655

267

1,024

38

557

33

6,103

Ac-227
Ag-108m
Am-241
Am-242
Am-242m
Am-243
C-14
Cd-113m
Cl-36
Cm-242
Cm-243
Cm-244
Cm-245
Cm-246
Cs-135
Cs-137
Eu-152
H-3
Ho-166m
I-129
Mo-93
Nb-93m
Nb-94
Ni-59
Ni-63
Np-237
Np-238
Np-239Pa-231
Pa-233
Pa-234m
Pb-210
Pd-107
Pu-238
Pu-239Pu-240Pu-241
Pu-242
Ra-226
Se-79
Sm-151
Sn-121m
Sn-126
Sr-90
Tc-99
Th-229Th-230Th-232
Th-234
U-233
U-234
U-235
U-236
U-237
U-238
Zr-93

3.52E+09
1.96E+12
6.61E+17
1.64E+15
1.65E+15
5.28E+15
2.22E+14
1.55E+13
1.02E+12
1.36E+15
1.86E+15
1.87E+17
7.06E+13
1.43E+13
9.83E+13
9.30E+18
2.09E+14
1.51E+16
1.80E+13
5.61E+12
7.99E+11
5.41E+15
1.99E+13
7.99E+14
8.50E+16
6.99E+13
7.43E+12
5.28E+15
5.90E+09
6.99E+13
5.39E+13
1.83E+08
2.37E+13
5.25E+17
5.56E+16
1.03E+17
3.97E+18
4.38E+14
5.93E+08
1.50E+13
5.45E+16
4.97E+15
1.10E+14
6.20E+18
2.77E+15
6.10E+07
7.18E+10
1.07E+05
5.39E+13
1.95E+10
2.25E+14
2.66E+12
5.49E+13
9.50E+13
5.39E+13
4.24E+14

6.47E+08
3.38E+11
1.12E+17
2.90E+14
2.92E+14
1.22E+15
3.72E+13
1.42E+12
1.88E+11
2.40E+14
3.11E+14
3.51E+16
2.24E+13
6.16E+12
1.71E+13
1.22E+18
1.90E+13
1.42E+15
3.44E+12
9.83E+11
1.39E+11
5.94E+14
3.39E+12
1.34E+14
1.35E+16
1.24E+13
1.31E+12
1.22E+15
9.21E+08
1.24E+13
7.79E+12
5.10E+07
4.48E+12
9.97E+16
8.00E+15
1.72E+16
3.66E+17
8.97E+13
1.41E+08
2.53E+12
7.53E+15
7.35E+14
1.96E+13
7.60E+17
4.61E+14
1.67E+07
1.38E+10
2.12E+04
7.79E+12
3.72E+09
3.69E+13
2.61E+11
8.35E+12
8.76E+12
7.79E+12
7.10E+13

1.87E+09
1.32E+12
3.87E+17
1.22E+15
1.22E+15
4.71E+15
1.52E+14
1.25E+13
7.26E+11
1.01E+15
1.81E+15
2.57E+17
8.69E+13
2.39E+13
6.61E+13
6.93E+18
1.75E+14
1.38E+16
1.33E+13
3.80E+12
5.40E+11
4.41E+15
1.31E+13
5.17E+14
5.85E+16
4.56E+13
5.51E+12
4.71E+15
3.21E+09
4.56E+13
3.01E+13
6.61E+07
1.73E+13
4.40E+17
3.10E+16
6.63E+16
3.17E+18
3.47E+14
2.36E+08
9.78E+12
3.31E+16
3.55E+15
7.59E+13
4.43E+18
1.78E+15
5.07E+07
3.29E+10
5.53E+04
3.01E+13
1.10E+10
1.23E+14
1.01E+12
3.22E+13
7.59E+13
3.01E+13
2.74E+14

4.48E+08
2.19E+11
1.20E+17
6.20E+14
6.23E+14
9.30E+14
2.55E+13
2.17E+12
1.15E+11
5.13E+14
3.34E+14
3.07E+16
3.85E+13
5.53E+12
1.36E+13
9.29E+17
2.47E+13
1.23E+15
1.99E+12
6.71E+11
9.24E+10
4.94E+14
2.32E+12
9.38E+13
9.49E+15
8.15E+12
2.80E+12
9.30E+14
7.20E+08
8.15E+12
6.51E+12
3.25E+07
3.00E+12
6.92E+16
8.82E+15
1.72E+16
4.50E+17
6.85E+13
9.65E+07
1.72E+12
7.47E+15
5.39E+14
1.32E+13
5.95E+17
3.22E+14
7.56E+06
1.04E+10
1.33E+04
6.51E+12
2.43E+09
2.95E+13
3.42E+11
5.99E+12
1.08E+13
6.51E+12
4.83E+13

1.69E+09
4.66E+16
3.34E+17
6.17E+14
6.20E+14
2.73E+15
7.26E+13
2.93E+15
3.37E+11
5.11E+14
8.40E+14
9.44E+16
3.94E+13
7.45E+12
3.79E+13
4.19E+18
7.07E+13
6.48E+15
8.62E+12
2.56E+12
1.22E+12
1.56E+17
6.07E+14
1.92E+14
2.02E+16
3.52E+13
2.79E+12
2.73E+15
2.80E+09
3.52E+13
2.19E+13
1.12E+08
1.10E+13
2.51E+17
2.60E+16
4.18E+16
1.92E+18
2.15E+14
3.59E+08
6.84E+12
2.69E+16
9.00E+14
5.01E+13
2.78E+18
1.25E+15
2.85E+07
4.28E+10
4.91E+04
2.19E+13
9.33E+09
1.30E+14
1.24E+12
2.50E+13
4.59E+13
2.19E+13
1.85E+14

8.52E+07
1.68E+15
1.49E+16
2.81E+13
2.82E+13
1.93E+14
3.28E+12
4.71E+13
1.66E+10
2.32E+13
3.81E+13
4.86E+15
4.51E+12
1.30E+12
1.82E+12
1.32E+17
1.25E+12
1.22E+14
4.34E+11
1.24E+11
5.21E+10
3.27E+15
2.59E+13
8.07E+12
7.66E+14
1.72E+12
1.27E+11
1.93E+14
1.14E+08
1.72E+12
8.05E+11
1.00E+07
5.92E+11
1.34E+16
9.62E+14
1.88E+15
3.51E+16
1.25E+13
2.58E+07
3.12E+11
9.45E+14
3.51E+13
2.50E+12
7.96E+16
5.62E+13
2.42E+06
2.29E+09
2.75E+03
8.05E+11
5.21E+08
5.55E+12
2.61E+10
9.71E+11
8.39E+11
8.05E+11
8.27E+12

7.55E+08
2.61E+16
1.47E+17
3.32E+14
3.33E+14
2.12E+15
3.65E+13
3.83E+15
1.82E+11
2.75E+14
6.32E+14
1.09E+17
4.96E+13
1.43E+13
2.01E+13
2.14E+18
3.92E+13
4.30E+15
4.82E+12
1.37E+12
5.92E+11
8.07E+16
2.85E+14
9.10E+13
9.65E+15
1.82E+13
1.50E+12
2.12E+15
1.16E+09
1.82E+13
8.84E+12
6.31E+07
6.51E+12
1.66E+17
1.06E+16
2.06E+16
1.00E+18
1.38E+14
1.75E+08
3.43E+12
1.17E+16
4.71E+14
2.75E+13
1.33E+18
6.18E+14
2.19E+07
1.81E+10
2.28E+04
8.84E+12
4.61E+09
5.43E+13
2.87E+11
1.07E+13
2.40E+13
8.84E+12
9.10E+13

3.69E+07
1.52E+15
1.21E+16
3.07E+13
3.09E+13
1.17E+14
2.20E+12
1.29E+14
9.60E+09
2.54E+13
3.24E+13
4.00E+15
2.23E+12
3.99E+11
1.15E+12
1.27E+17
2.65E+12
2.33E+14
2.47E+11
7.48E+10
3.80E+10
4.43E+15
1.90E+13
6.04E+12
6.19E+14
9.38E+11
1.39E+11
1.17E+14
6.64E+07
9.38E+11
6.35E+11
1.96E+06
3.50E+11
7.95E+15
8.46E+14
1.60E+15
7.22E+16
8.61E+12
6.95E+06
1.88E+11
8.21E+14
2.75E+13
1.49E+12
8.22E+16
3.50E+13
6.41E+05
9.53E+08
1.05E+03
6.35E+11
2.33E+08
3.37E+12
3.20E+10
6.20E+11
1.73E+12
6.35E+11
5.04E+12

9.06E+09
7.59E+16
1.79E+18
4.78E+15
4.80E+15
1.73E+16
5.52E+14
6.98E+15
2.59E+12
3.95E+15
5.86E+15
7.22E+17
3.14E+14
7.33E+13
2.56E+14
2.50E+19
5.42E+14
4.27E+16
5.09E+13
1.52E+13
3.47E+12
2.55E+17
9.76E+14
1.84E+15
1.98E+17
1.92E+14
2.16E+13
1.73E+16
1.49E+10
1.92E+14
1.31E+14
5.20E+08
6.70E+13
1.57E+18
1.42E+17
2.70E+17
1.10E+19
1.32E+15
1.63E+09
3.98E+13
1.43E+17
1.12E+16
3.00E+14
1.63E+19
7.30E+15
1.89E+08
1.93E+11
2.72E+05
1.31E+14
5.13E+10
6.08E+14
5.86E+12
1.39E+14
2.63E+14
1.31E+14
1.11E+15

TR-10-13

77

Table C-6. Radionuclide inventory in BWR I type canister (12 assemblies, burnup
40.4 MWd/kgHM, age 37 years. Inventory for fuel matrix based on information in /­SKBdoc
1221579/ and construction material, Crud and PWR control rod estimated from data givenin /­SKBdoc 1179234/.
Radionuclide

Ac-227
Ag-108m
Am-241
Am-242
Am-242m
Am-243
C-14
Cd-113m
Cl-36
Cm-242
Cm-243
Cm-244
Cm-245
Cm-246
Cs-135
Cs-137
Eu-152
H-3
Ho-166m
I-129
Mo-93
Nb-93m
Nb-94
Ni-59
Ni-63
Np-237
Np-238
Np-239Pa-231
Pa-233
Pa-234m
Pb-210
Pd-107
Pu-238
Pu-239Pu-240Pu-241
Pu-242
Ra-226
Se-79
Sm-151
Sn-121m
Sn-126
Sr-90
Tc-99
Th-229Th-230Th-232
Th-234
U-233
U-234
U-235
U-236
U–237
U-238
Zr-93

78

BWR I – Radionuclide inventory (Bq/canister)Constr. mtrl.
Crud
UO2-matrix

Total

1.60E+06
4.99E+08
2.99E+14
7.45E+11
7.48E+11
2.39E+12
2.84E+10
7.04E+09
4.54E+08
6.16E+11
8.44E+11
8.45E+13
3.20E+10
6.46E+09
4.45E+10
4.21E+15
9.45E+10
6.85E+12
8.16E+09
2.54E+09
6.49E+07
1.36E+11
1.16E+07
2.46E+08
2.58E+10
3.17E+10
3.37E+09
2.39E+12
2.67E+06
3.17E+10
2.44E+10
8.30E+04
1.08E+10
2.38E+14
2.52E+13
4.67E+13
1.80E+15
1.98E+11
2.69E+05
6.81E+09
2.47E+13
7.65E+11
4.96E+10
2.81E+15
1.26E+12
2.76E+04
3.25E+07
4.84E+01
2.44E+10
6.31E+06
1.02E+11
1.21E+09
2.49E+10
4.30E+10
2.44E+10
1.66E+11

1.60E+06
8.89E+08
2.99E+14
7.45E+11
7.48E+11
2.39E+12
1.01E+11
7.04E+09
4.62E+08
6.16E+11
8.44E+11
8.45E+13
3.20E+10
6.46E+09
4.45E+10
4.21E+15
9.45E+10
6.85E+12
8.16E+09
2.54E+09
3.62E+08
2.45E+12
9.00E+09
3.62E+11
3.85E+13
3.17E+10
3.37E+09
2.39E+12
2.67E+06
3.17E+10
2.44E+10
8.30E+04
1.08E+10
2.38E+14
2.52E+13
4.67E+13
1.80E+15
1.98E+11
2.69E+05
6.81E+09
2.47E+13
2.25E+12
4.96E+10
2.81E+15
1.26E+12
2.76E+04
3.25E+07
4.84E+01
2.44E+10
8.84E+06
1.02E+11
1.21E+09
2.49E+10
4.30E+10
2.44E+10
1.92E+11

0.00E+00

1.30E–01
3.90E+08
2.70E+07
1.27E+05
1.27E+05
2.90E+05

7.23E+10
0.00E+00
7.10E+06
1.05E+05
7.39E+04
1.22E+07
5.38E+03
1.39E+03
1.28E+09
0.00E+00
0.00E+00
0.00E+00
2.92E+08
2.26E+12
8.79E+09
3.57E+11
3.79E+13

0.00E+00

2.30E+08

0.00E+00
0.00E+00
1.48E+12
1.20E+09
5.03E+07

2.53E+06

2.58E+10

5.39E+06
5.23E+10
2.00E+08
5.05E+09
5.40E+11
2.77E+03
5.72E+02
2.90E+05
2.14E–01
2.77E+03
2.00E+03
7.34E–03
2.43E+07
1.96E+06
3.10E+06
1.56E+08
1.96E+04
2.33E–02

1.21E+08

8.41E+05
3.11E–03
2.77E+00
3.80E–06
2.00E+03
5.45E–01
8.74E+03
9.13E+01
1.88E+03
3.73E+03
2.00E+03
2.18E+06

TR-10-13

2,208 BWR I canisters:•

UO2-values from SKBdoc 1221579 (‘Nuklidlista 100320.xls’)•

Inventory for construction material from SKBdoc 1179234 by linear interpolation between 38 and
60 MWd/kg U to find inventory at 40.4 MWd/kg U.
(Data from the files ‘88_Ind-B38-000.xls’ and ‘89_Ind-B60-000.xls’)•

Inventory for Crud from SKBdoc 1179234 by linear interpolation between 38 and 60 MWd/kg U
to find inventory at 40.4 MWd/kg U.
(Data from the files ‘95_Crud-B38-000.xls’ and ‘96_Crud-B60-000.xls’)

TR-10-13

79

Table C-7. Radionuclide inventory in PWR I type canister (4 assemblies, burnup 44.8 MWd/kgHM,
age 38 years). Inventory for fuel matrix based on information in /­SKBdoc 1221579/ and construction material and Crud estimated from data given in /­SKBdoc 1179234/.
Radionuclide

Ac-227
Ag-108m
Am-241
Am-242
Am-242m
Am-243
C-14
Cd-113m
Cl-36
Cm-242
Cm-243
Cm-244
Cm-245
Cm-246
Cs-135
Cs-137
Eu-152
H-3
Ho-166m
I-129
Mo-93
Nb-93m
Nb-94
Ni-59
Ni-63
Np-237
Np-238
Np-239Pa-231
Pa-233
Pa-234m
Pb-210
Pd-107
Pu-238
Pu-239Pu-240Pu-241
Pu-242
Ra-226
Se-79
Sm-151
Sn-121m
Sn-126
Sr-90
Tc-99
Th-229Th-230Th-232
Th-234
U-233
U-234
U-235
U-236
U-237
U-238
Zr-93

80

PWR I – Radionuclide inventory (Bq/canister)Constr. mtrl.
Crud
UO2-matrix

1.65E+06
3.24E+08
3.26E+14
6.03E+11
6.06E+11
2.67E+12
2.26E+10
7.16E+09
3.23E+08
4.99E+11
8.20E+11
9.22E+13
3.85E+10
7.27E+09
3.71E+10
4.10E+15
6.90E+10
6.32E+12
8.42E+09
2.50E+09
8.22E+07
1.33E+11
1.15E+07
1.74E+08
1.80E+10
3.44E+10
2.73E+09
2.67E+12
2.74E+06
3.44E+10
2.14E+10
1.09E+05
1.07E+10
2.45E+14
2.54E+13
4.08E+13
1.87E+15
2.10E+11
3.51E+05
6.68E+09
2.62E+13
7.50E+11
4.90E+10
2.72E+15
1.23E+12
2.79E+04
4.18E+07
4.79E+01
2.14E+10
7.41E+06
1.27E+11
1.21E+09
2.44E+10
4.48E+10
2.14E+10
1.63E+11

2.22E+08

1.85E–01
5.53E+07
4.42E+07
1.41E+05
1.42E+05
4.33E+05

4.59E+10
0.00E+00
5.34E+06

Control rod

4.55E+13

2.49E+09
2.86E+12
7.47E+05
1.17E+05
1.13E+05
1.92E+07
1.01E+04
2.63E+03

7.18E+08
0.00E+00
1.35E+10
0.00E+00
9.86E+08
1.52E+14
5.93E+11
1.71E+11
1.82E+13

0.00E+00

1.82E+08

3.24E+06
0.00E+00
1.29E+11
6.73E+08
1.47E+08

1.70E+06

1.74E+10

0.00E+00
0.00E+00
9.76E+08
0.00E+00
6.78E+04
1.09E+08
4.14E+05
3.18E+08
3.06E+10
4.84E+03
6.37E+02
4.33E+05
3.14E–01
4.84E+03
2.94E+03
1.23E–02
3.97E+07
3.47E+06
5.47E+06
2.50E+08
2.93E+04
3.96E–02

3.55E+05

1.16E+04
4.41E–03
4.78E+00
6.42E–06
2.94E+03
9.99E–01
1.52E+04
1.63E+02
3.21E+03
5.99E+03
2.94E+03
1.31E+04

1.22E+08
3.38E+10
2.14E+08
1.63E+10
1.43E+12

0.00E+00

0.00E+00

1.17E+06
0.00E+00
3.19E+07
0.00E+00
1.93E+07

0.00E+00

1.66E+03

Total

1.65E+06
4.55E+13
3.26E+14
6.03E+11
6.06E+11
2.67E+12
7.09E+10
2.86E+12
3.29E+08
4.99E+11
8.20E+11
9.22E+13
3.85E+10
7.27E+09
3.71E+10
4.10E+15
6.90E+10
6.33E+12
8.42E+09
2.50E+09
1.19E+09
1.52E+14
5.93E+11
1.88E+11
1.97E+13
3.44E+10
2.73E+09
2.67E+12
2.74E+06
3.44E+10
2.14E+10
1.09E+05
1.07E+10
2.45E+14
2.54E+13
4.08E+13
1.87E+15
2.10E+11
3.51E+05
6.68E+09
2.62E+13
8.79E+11
4.90E+10
2.72E+15
1.23E+12
2.79E+04
4.18E+07
4.79E+01
2.14E+10
9.11E+06
1.27E+11
1.21E+09
2.44E+10
4.48E+10
2.14E+10
1.80E+11

TR-10-13

1,024 PWR I canisters:•

UO2-values from SKBdoc 1221579 (‘Nuklidlista 100320.xls’)•

Inventory for construction material from SKBdoc 1179234 by linear interpolation between 30 and
60 MWd/kg U to find inventory at 44.8 MWd/kg U.
(Data from the files ‘90_Ind-P30-000.xls’ and ‘91_Ind-P60-000.xls’)•

Inventory for Crud from SKBdoc 1179234 by linear interpolation between 30 and 60 MWd/kg U
to find inventory at 44.8 MWd/kg U.
(Data from the files ‘97_Crud-P30-000.xls’ and ‘98_Crud-P60-000.xls’)•

Inventory for control rods from SKBdoc 1179234 (‘110_PWR-ss.xls’). One control rod cluster
per canister is assumed.

TR-10-13

81

Table C-8. Radionuclide inventory in BWR II type canister (12 assemblies, burnup
47.8 MWd/kgHM, age 48 years). Inventory for fuel matrix based on information in /­SKBdoc
1221579/ and construction material and Crud estimated from data given in /­SKBdoc 1179234/.
Radionuclide

Ac-227
Ag-108m
Am-241
Am-242
Am-242m
Am-243
C-14
Cd-113m
Cl-36
Cm-242
Cm-243
Cm-244
Cm-245
Cm-246
Cs-135
Cs-137
Eu-152
H-3
Ho-166m
I-129
Mo-93
Nb-93m
Nb-94
Ni-59
Ni-63
Np-237
Np-238
Np-239Pa-231
Pa-233
Pa-234m
Pb-210
Pd-107
Pu-238
Pu-239Pu-240Pu-241
Pu-242
Ra-226
Se-79
Sm-151
Sn-121m
Sn-126
Sr-90
Tc-99
Th-229Th-230Th-232
Th-234
U-233
U-234
U-235
U-236
U-237
U-238
Zr-93

82

BWR II – Radionuclide inventory (Bq/canister)Constr. mtrl.
Crud
UO2-matrix

Total

2.02E+06
5.58E+08
3.49E+14
9.04E+11
9.08E+11
3.79E+12
3.01E+10
4.43E+09
5.76E+08
7.48E+11
9.70E+11
1.09E+14
6.99E+10
1.92E+10
5.33E+10
3.80E+15
5.93E+10
4.42E+12
1.07E+10
3.06E+09
8.01E+07
1.69E+11
1.48E+07
3.04E+08
3.03E+10
3.85E+10
4.09E+09
3.79E+12
2.87E+06
3.85E+10
2.43E+10
1.59E+05
1.40E+10
3.11E+14
2.49E+13
5.36E+13
1.14E+15
2.79E+11
4.38E+05
7.88E+09
2.35E+13
8.09E+11
6.11E+10
2.37E+15
1.44E+12
5.20E+04
4.30E+07
6.59E+01
2.43E+10
9.01E+06
1.15E+11
8.15E+08
2.60E+10
2.73E+10
2.43E+10
1.91E+11

2.02E+06
1.05E+09
3.49E+14
9.04E+11
9.08E+11
3.79E+12
1.16E+11
4.43E+09
5.84E+08
7.48E+11
9.70E+11
1.09E+14
6.99E+10
1.92E+10
5.33E+10
3.80E+15
5.93E+10
4.42E+12
1.07E+10
3.06E+09
4.34E+08
1.85E+12
1.06E+10
4.16E+11
4.20E+13
3.85E+10
4.09E+09
3.79E+12
2.87E+06
3.85E+10
2.43E+10
1.59E+05
1.40E+10
3.11E+14
2.49E+13
5.36E+13
1.14E+15
2.79E+11
4.38E+05
7.88E+09
2.35E+13
2.29E+12
6.11E+10
2.37E+15
1.44E+12
5.20E+04
4.30E+07
6.59E+01
2.43E+10
1.16E+07
1.15E+11
8.15E+08
2.60E+10
2.73E+10
2.43E+10
2.21E+11

0.00E+00

1.84E–01
4.96E+08
3.12E+07
1.50E+05
1.51E+05
3.91E+05

8.59E+10
0.00E+00
8.43E+06
1.24E+05
7.60E+04
1.30E+07
1.02E+04
3.08E+03
1.16E+09
0.00E+00
0.00E+00
0.00E+00
3.46E+08
1.63E+12
1.03E+10
4.09E+11
4.12E+13

0.00E+00

2.31E+08

0.00E+00
0.00E+00
1.48E+12
1.08E+09
5.89E+07

2.56E+06

3.05E+10

7.79E+06
4.59E+10
2.84E+08
7.03E+09
7.14E+11
3.53E+03
6.77E+02
3.91E+05
2.66E–01
3.53E+03
1.98E+03
1.48E–02
3.16E+07
2.05E+06
3.38E+06
9.81E+07
2.30E+04
4.04E–02

1.46E+08

1.18E+06
5.96E–03
3.94E+00
5.69E–06
1.98E+03
8.31E–01
1.07E+04
8.81E+01
2.18E+03
2.35E+03
1.98E+03
3.15E+06

TR-10-13

321 BWR II canisters:•

UO2-values from SKBdoc 1221579 (‘Nuklidlista 100320.xls’)•

Inventory for construction material from SKBdoc 1179234 by linear interpolation between 38 and
60 MWd/kg U to find inventory at 47.8 MWd/kg U.
(Data from the files ‘88_Ind-B38-000.xls’ and ‘89_Ind-B60-000.xls’)•

Inventory for Crud from SKBdoc 1179234 by linear interpolation between 38 and 60 MWd/kg U
to find inventory at 47.8 MWd/kg U.
(Data from the files ‘95_Crud-B38-000.xls’ and ‘96_Crud-B60-000.xls’)

TR-10-13

83

Table C-9. Radionuclide inventory in BWR III type canister (9 assemblies, burnup
47.8 MWd/kgHM, age 32 years). Inventory for fuel matrix based on information in /­SKBdoc
1221579/ and construction material and Crud estimated from data given in /­SKBdoc 1179234/.
Radionuclide

Ac-227
Ag-108m
Am-241
Am-242
Am-242m
Am-243
C-14
Cd-113m
Cl-36
Cm-242
Cm-243
Cm-244
Cm-245
Cm-246
Cs-135
Cs-137
Eu-152
H-3
Ho-166m
I-129
Mo-93
Nb-93m
Nb-94
Ni-59
Ni-63
Np-237
Np-238
Np-239Pa-231
Pa-233
Pa-234m
Pb-210
Pd-107
Pu-238
Pu-239Pu-240Pu-241
Pu-242
Ra-226
Se-79
Sm-151
Sn-121m
Sn-126
Sr-90
Tc-99
Th-229Th-230Th-232
Th-234
U-233
U-234
U-235
U-236
U-237
U-238
Zr-93

84

BWR III – Radionuclide inventory (Bq/canister)Constr. mtrl.
Crud
UO2-matrix

Total

1.13E+06
4.18E+08
2.34E+14
7.36E+11
7.39E+11
2.85E+12
2.74E+10
7.55E+09
4.32E+08
6.09E+11
1.09E+12
1.55E+14
5.25E+10
1.44E+10
4.00E+10
4.19E+15
1.06E+11
8.36E+12
8.05E+09
2.30E+09
6.01E+07
1.10E+11
1.11E+07
2.28E+08
2.26E+10
2.75E+10
3.33E+09
2.85E+12
1.94E+06
2.75E+10
1.82E+10
4.00E+04
1.05E+10
2.66E+14
1.87E+13
4.01E+13
1.92E+15
2.10E+11
1.43E+05
5.91E+09
2.00E+13
7.49E+11
4.58E+10
2.68E+15
1.08E+12
3.06E+04
1.99E+07
3.34E+01
1.82E+10
4.71E+06
7.46E+10
6.11E+08
1.95E+10
4.59E+10
1.82E+10
1.43E+11

1.13E+06
8.00E+08
2.34E+14
7.36E+11
7.39E+11
2.85E+12
9.20E+10
7.55E+09
4.38E+08
6.09E+11
1.09E+12
1.55E+14
5.25E+10
1.44E+10
4.00E+10
4.19E+15
1.06E+11
8.36E+12
8.05E+09
2.30E+09
3.26E+08
2.66E+12
7.92E+09
3.12E+11
3.54E+13
2.75E+10
3.33E+09
2.85E+12
1.94E+06
2.75E+10
1.82E+10
4.00E+04
1.05E+10
2.66E+14
1.87E+13
4.01E+13
1.92E+15
2.10E+11
1.43E+05
5.91E+09
2.00E+13
2.15E+12
4.58E+10
2.68E+15
1.08E+12
3.06E+04
1.99E+07
3.34E+01
1.82E+10
6.63E+06
7.46E+10
6.11E+08
1.95E+10
4.59E+10
1.82E+10
1.66E+11

0.00E+00

1.02E–01
3.82E+08
2.10E+07
1.22E+05
1.23E+05
2.94E+05

6.46E+10
0.00E+00
6.32E+06
1.01E+05
8.56E+04
1.85E+07
7.69E+03
2.32E+03
1.28E+09
0.00E+00
0.00E+00
0.00E+00
2.60E+08
2.48E+12
7.70E+09
3.07E+11
3.47E+13

0.00E+00

1.74E+08

0.00E+00
0.00E+00
1.40E+12
1.34E+08
4.42E+07

1.92E+06

2.29E+10

5.87E+06
6.98E+10
2.13E+08
5.28E+09
6.02E+11
2.52E+03
5.51E+02
2.94E+05
1.77E–01
2.52E+03
1.48E+03
3.81E–03
2.70E+07
1.54E+06
2.51E+06
1.65E+08
1.72E+04
1.33E–02

1.38E+08

8.88E+05
3.70E–03
1.82E+00
2.92E–06
1.48E+03
4.36E–01
6.83E+03
6.60E+01
1.63E+03
3.94E+03
1.48E+03
2.36E+06

TR-10-13

1,655 BWR III canisters:•

UO2-values from SKBdoc 1221579 (‘Nuklidlista 100320.xls’)•

Inventory for construction material from SKBdoc 1179234 by linear interpolation between 38 and
60 MWd/kg U to find inventory at 47.8 MWd/kg U.
(Data from the files ‘88_Ind-B38-000.xls’ and ‘89_Ind-B60-000.xls’)•

Inventory for Crud from SKBdoc 1179234 by linear interpolation between 38 and 60 MWd/kg U
to find inventory at 47.8 MWd/kg U.
(Data from the files ‘95_Crud-B38-000.xls’ and ‘96_Crud-B60-000.xls’)

TR-10-13

85

Table C-10. Radionuclide inventory in BWR MOX type canister (11 BWR assemblies, burnup
37.7 MWd/kgHM, age 43 years and 1 MOX assembly, burnup 50 MWd/kgHM, age 50 years).
Inventory for fuel matrix based on information in /­SKBdoc 1221579/ and construction material
and Crud estimated from data given in /­SKBdoc 1179234/.
Radionuclide

Ac-227

Ag-108m
Am-241
Am-242
Am-242m
Am-243
C-14
Cd-113m
Cl-36
Cm-242
Cm-243
Cm-244
Cm-245
Cm-246
Cs-135
Cs-137
Eu-152
H-3
Ho-166m
I-129
Mo-93
Nb-93m
Nb-94
Ni-59
Ni-63
Np-237
Np-238
Np-239Pa-231
Pa-233
Pa-234m
Pb-210
Pd-107
Pu-238
Pu-239Pu-240Pu-241
Pu-242
Ra-226
Se-79
Sm-151
Sn-121m
Sn-126
Sr-90
Tc-99
Th-229Th-230Th-232
Th-234
U-233
U-234
U-235
U-236
U-237
U-238
Zr-93

86

BWRMOX – Radionuclide inventory (Bq/canister)Constr. mtrl.
Crud
UO2-matrix

Total

1.68E+06
4.55E+08
4.48E+14
2.32E+12
2.33E+12
3.48E+12
2.59E+10
8.13E+09
4.23E+08
1.92E+12
1.25E+12
1.15E+14
1.44E+11
2.07E+10
5.08E+10
3.48E+15
9.26E+10
4.59E+12
7.44E+09
2.51E+09
6.06E+07
1.34E+11
1.20E+07
2.30E+08
2.29E+10
3.05E+10
1.05E+10
3.48E+12
2.70E+06
3.05E+10
2.44E+10
1.22E+05
1.12E+10
2.59E+14
3.30E+13
6.43E+13
1.69E+15
2.57E+11
3.62E+05
6.46E+09
2.80E+13
7.02E+11
4.94E+10
2.23E+15
1.21E+12
2.83E+04
3.90E+07
4.99E+01
2.44E+10
6.57E+06
1.11E+11
1.28E+09
2.24E+10
4.04E+10
2.44E+10
1.56E+11

1.68E+06
8.19E+08
4.48E+14
2.32E+12
2.33E+12
3.48E+12
9.55E+10
8.13E+09
4.30E+08
1.92E+12
1.25E+12
1.15E+14
1.44E+11
2.07E+10
5.08E+10
3.48E+15
9.26E+10
4.59E+12
7.44E+09
2.51E+09
3.46E+08
1.85E+128.70E+09
3.51E+11
3.56E+13
3.05E+10
1.05E+10
3.48E+12
2.70E+06
3.05E+10
2.44E+10
1.22E+05
1.12E+10
2.59E+14
3.30E+13
6.43E+13
1.69E+15
2.57E+11
3.62E+05
6.46E+09
2.80E+13
2.02E+12
4.94E+10
2.23E+15
1.21E+12
2.83E+04
3.90E+07
4.99E+01
2.44E+10
9.09E+06
1.11E+11
1.28E+09
2.24E+10
4.04E+10
2.44E+10
1.81E+11

0.00E+00

1.38E–01
3.64E+08
3.36E+07
1.78E+05
1.79E+05
3.60E+05

6.96E+10
0.00E+00
6.84E+06
1.47E+05
8.15E+04
1.25E+07
1.14E+04
3.23E+03
1.07E+09
0.00E+00
0.00E+00
0.00E+00
2.81E+08
1.68E+12
8.50E+09
3.46E+11
3.51E+13

0.00E+00

2.30E+08

0.00E+00
0.00E+00
1.32E+12
9.98E+08
4.86E+07

2.52E+06

2.48E+10

4.91E+06
3.64E+10
1.84E+08
4.66E+09
4.72E+11
2.55E+03
8.03E+02
3.60E+05
2.13E–01
2.55E+03
2.00E+03
1.11E–02
2.31E+07
2.12E+06
3.96E+06
1.29E+08
2.51E+04
3.21E–02

1.01E+08

7.74E+05
2.60E–03
3.35E+00
3.74E–06
2.00E+03
5.55E–01
9.35E+03
8.56E+01
1.64E+03
3.09E+03
2.00E+03
2.00E+06

TR-10-13

267 BWRMOX canisters:•

UO2-values from SKBdoc 1221579 (‘Nuklidlista 100320.xls’)•

Inventory for construction material from SKBdoc 1179234 at 38 MWd/kg U for 11 BWR
assemblies and 50 MWd/kg U for 1 MOX assembly.
(Data from the files ‘88_Ind-B38-000.xls’ and ‘93_Ind-M50-000.xls’)•

Inventory for Crud from SKBdoc 1179234 at 38 MWd/kg U for 11 BWR assemblies and
50 MWd/kg U for 1 MOX assembly.
(Data from the files ‘95_Crud-B38-000.xls’ and ‘100_Crud-M50-000.xls’)

TR-10-13

87

Table C-11. Radionuclide inventory in PWR II type canister (4 assemblies, burnup 57 MWd/kgHM,
age 55 years Inventory for fuel matrix based on information in /­SKBdoc 1221579/ and construction material and Crud estimated from data given in /­SKBdoc 1179234/.
Radionuclide

Ac-227
Ag-108m
Am-241
Am-242
Am-242m
Am-243
C-14
Cd-113m
Cl-36
Cm-242
Cm-243
Cm-244
Cm-245
Cm-246
Cs-135
Cs-137
Eu-152
H-3
Ho-166m
I-129
Mo-93
Nb-93m
Nb-94
Ni-59
Ni-63
Np-237
Np-238
Np-239Pa-231
Pa-233
Pa-234m
Pb-210
Pd-107
Pu-238
Pu-239Pu-240Pu-241
Pu-242
Ra-226
Se-79
Sm-151
Sn-121m
Sn-126
Sr-90
Tc-99
Th-229Th-230Th-232
Th-234
U-233
U-234
U-235
U-236
U-237
U-238
Zr-93

88

PWR II – Radionuclide inventory (Bq/canister)Constr. mtrl.
Crud
UO2-matrix

2.24E+06
3.54E+08
3.93E+14
7.39E+11
7.42E+11
5.07E+12
3.03E+10
3.62E+09
4.29E+08
6.11E+11
1.00E+12
1.28E+14
1.19E+11
3.42E+10
4.80E+10
3.48E+15
3.28E+10
3.21E+12
1.14E+10
3.27E+09
1.09E+08
1.81E+11
1.61E+07
2.23E+08
2.13E+10
4.53E+10
3.34E+09
5.07E+12
2.99E+06
4.53E+10
2.12E+10
2.63E+05
1.56E+10
3.53E+14
2.53E+13
4.96E+13
9.22E+14
3.30E+11
6.78E+05
8.21E+09
2.49E+13
8.06E+11
6.59E+10
2.10E+15
1.48E+12
6.38E+04
6.03E+07
7.24E+01
2.12E+10
1.20E+07
1.46E+11
6.88E+08
2.56E+10
2.21E+10
2.12E+10
1.97E+11

2.40E+08

2.97E–01
5.02E+07
5.62E+07
1.74E+05
1.75E+05
6.42E+05

5.35E+10
0.00E+00
6.24E+06

Control rod

4.43E+13

2.49E+09
1.24E+12
7.47E+05
1.44E+05
1.08E+05
1.63E+07
1.71E+04
4.60E+03

5.65E+08
0.00E+00
5.99E+09
0.00E+00
1.14E+09
8.59E+13
6.81E+11
1.95E+11
1.88E+13

0.00E+00

1.83E+08

3.78E+06
0.00E+00
1.18E+11
5.23E+08
1.66E+08

1.73E+06

2.03E+10

0.00E+00
0.00E+00
3.74E+08
0.00E+00
7.13E+04
5.62E+07
4.36E+05
3.37E+08
2.90E+10
6.75E+03
7.88E+02
6.42E+05
4.23E–01
6.75E+03
2.89E+03
3.69E–02
5.15E+07
3.74E+06
6.38E+06
1.29E+08
3.92E+04
9.53E–02

2.96E+05

1.22E+04
8.86E–03
8.53E+00
1.14E–05
2.89E+03
1.81E+00
2.08E+04
1.54E+02
3.99E+03
3.08E+03
2.89E+03
1.38E+04

1.21E+08
1.65E+10
2.13E+08
1.63E+10
1.27E+12

0.00E+00

0.00E+00

1.17E+06
0.00E+00
2.52E+07
0.00E+00
1.93E+07

0.00E+00

1.66E+03

Total

2.24E+06
4.43E+13
3.93E+14
7.39E+11
7.42E+11
5.07E+12
8.64E+10
1.24E+12
4.36E+08
6.11E+11
1.00E+12
1.28E+14
1.19E+11
3.42E+10
4.80E+10
3.48E+15
3.28E+10
3.22E+12
1.14E+10
3.27E+09
1.37E+09
8.61E+13
6.81E+11
2.12E+11
2.02E+13
4.53E+10
3.34E+09
5.07E+12
2.99E+06
4.53E+10
2.12E+10
2.63E+05
1.56E+10
3.53E+14
2.53E+13
4.96E+13
9.22E+14
3.30E+11
6.78E+05
8.22E+09
2.49E+13
9.23E+11
6.59E+10
2.10E+15
1.48E+12
6.38E+04
6.03E+07
7.24E+01
2.12E+10
1.37E+07
1.46E+11
6.88E+08
2.56E+10
2.21E+10
2.12E+10
2.18E+11

TR-10-13

38 PWR II canisters:•

UO2-values from SKBdoc 1221579 (‘Nuklidlista 100320.xls’)•

Inventory for construction material from SKBdoc 1179234 by linear interpolation between 30 and
60 MWd/kg U to find inventory at 57 MWd/kg U.
(Data from the files ‘90_Ind-P30-000.xls’ and ‘91_Ind-P60-000.xls’)•

Inventory for Crud from SKBdoc 1179234 by linear interpolation between 30 and 60 MWd/kg U
to find inventory at 57 MWd/kg U.
(Data from the files ‘97_Crud-P30-000.xls’ and ‘98_Crud-P60-000.xls’)•

Inventory for control rods from SKBdoc 1179234 (‘110_PWR-ss.xls’). One control rod cluster
per canister is assumed.

TR-10-13

89

Table C-12. Radionuclide inventory in PWR III type canister (2 assemblies, burnup 57 MWd/kgHM,
age 51 years and 1 assembly burnup 57 MWd/kgHM, age 20 years). Inventory for fuel matrix
based on information in /­SKBdoc 1221579/ and construction material and Crud estimated fromdata given in /­SKBdoc 1179234/.
Radionuclide

Ac-227
Ag-108m
Am-241
Am-242
Am-242m
Am-243
C-14
Cd-113m
Cl-36
Cm-242
Cm-243
Cm-244
Cm-245
Cm-246
Cs-135
Cs-137
Eu-152
H-3
Ho-166m
I-129
Mo-93
Nb-93m
Nb-94
Ni-59
Ni-63
Np-237
Np-238
Np-239Pa-231
Pa-233
Pa-234m
Pb-210
Pd-107
Pu-238
Pu-239Pu-240Pu-241
Pu-242
Ra-226
Se-79
Sm-151
Sn-121m
Sn-126
Sr-90
Tc-99
Th-229Th-230Th-232
Th-234
U-233
U-234
U-235
U-236
U-237
U-238
Zr-93

90

PWR III – Radionuclide inventory (Bq/canister)Constr. mtrl.
Crud
UO2-matrix

1.36E+06
2.88E+08
2.64E+14
5.96E+11
5.99E+11
3.81E+12
2.28E+10
7.22E+09
3.22E+08
4.93E+11
1.14E+12
1.96E+14
8.91E+10
2.57E+10
3.60E+10
3.85E+15
7.04E+10
7.71E+12
8.65E+09
2.45E+09
8.20E+07
1.19E+11
1.21E+07
1.68E+08
1.78E+10
3.28E+10
2.69E+09
3.81E+12
2.09E+06
3.28E+10
1.59E+10
1.13E+05
1.17E+10
2.98E+14
1.90E+13
3.70E+13
1.80E+15
2.47E+11
3.15E+05
6.16E+09
2.09E+13
7.36E+11
4.94E+10
2.39E+15
1.11E+12
3.93E+04
3.25E+07
4.09E+01
1.59E+10
6.98E+06
9.75E+10
5.16E+08
1.91E+10
4.31E+10
1.59E+10
1.48E+11

1.85E+08

1.75E–01
3.85E+07
3.79E+07
1.41E+05
1.41E+05
4.82E+05

4.02E+10
0.00E+00
4.68E+06

Control rod

4.69E+13

2.50E+09
6.88E+12
7.47E+05
1.16E+05
1.22E+05
2.49E+07
1.29E+04
3.46E+03

6.26E+08
0.00E+00
1.44E+10
0.00E+00
8.58E+08
1.45E+14
5.11E+11
1.47E+11
1.57E+13

0.00E+00

1.38E+08

2.83E+06
0.00E+00
1.10E+11
5.88E+08
1.25E+08

1.29E+06

1.52E+10

0.00E+00
0.00E+00
2.66E+09
0.00E+00
5.36E+04
9.47E+07
3.27E+05
2.53E+08
2.41E+10
4.89E+03
6.36E+02
4.82E+05
2.82E–01
4.89E+03
2.17E+03
1.59E–02
4.35E+07
2.80E+06
4.76E+06
2.52E+08
2.94E+04
4.42E–02

2.76E+05

9.14E+03
5.34E–03
4.58E+00
6.42E–06
2.17E+03
1.05E+00
1.39E+04
1.16E+02
2.99E+03
6.02E+03
2.17E+03
1.04E+04

1.22E+08
7.20E+10
2.14E+08
1.63E+10
1.62E+12

0.00E+00

0.00E+00

1.17E+06
0.00E+00
4.09E+07
0.00E+00
1.93E+07

0.00E+00

1.66E+03

Total

1.36E+06
4.69E+13
2.64E+14
5.96E+11
5.99E+11
3.81E+12
6.55E+10
6.88E+12
3.27E+08
4.93E+11
1.14E+12
1.96E+14
8.91E+10
2.57E+10
3.60E+10
3.85E+15
7.04E+10
7.73E+12
8.65E+09
2.45E+09
1.06E+09
1.45E+14
5.11E+11
1.63E+11
1.73E+13
3.28E+10
2.69E+09
3.81E+12
2.09E+06
3.28E+10
1.59E+10
1.13E+05
1.17E+10
2.98E+14
1.90E+13
3.70E+13
1.80E+15
2.47E+11
3.15E+05
6.16E+09
2.09E+13
8.46E+11
4.94E+10
2.39E+15
1.11E+12
3.93E+04
3.25E+07
4.09E+01
1.59E+10
8.27E+06
9.75E+10
5.16E+08
1.91E+10
4.31E+10
1.59E+10
1.63E+11

TR-10-13

557 PWR III canisters:•

UO2-values from SKBdoc 1221579 (‘Nuklidlista 100320.xls’)•

Inventory for construction material from SKBdoc 1179234 by linear interpolation between 30 and
60 MWd/kg U to find inventory at 57 MWd/kg U.
(Data from the files ‘90_Ind-P30-000.xls’ and ‘91_Ind-P60-000.xls’)•

Inventory for Crud from SKBdoc 1179234 by linear interpolation between 30 and 60 MWd/kg U
to find inventory at 57 MWd/kg U.
(Data from the files ‘97_Crud-P30-000.xls’ and ‘98_Crud-P60-000.xls’)•

Inventory for control rods from SKBdoc 1179234 (‘110_PWR-ss.xls’). One control rod cluster
per canister is assumed.

TR-10-13

91

Table C-13. Radionuclide inventory in PWR MOX type canister (3 PWR assemblies, burnup
44.8 MWd/kgHM, age 32 years and 1 MOX assembly burnup 34.8 MWd/kgHM, age 57 years).
Inventory for fuel matrix based on information in /­SKBdoc 1221579/ and construction material
and Crud estimated from data given in /­SKBdoc 1179234/.
Radionuclide

Ac-227
Ag-108m
Am-241
Am-242
Am-242m
Am-243
C-14
Cd-113m
Cl-36
Cm-242
Cm-243
Cm-244
Cm-245
Cm-246
Cs-135
Cs-137
Eu-152
H-3
Ho-166m
I-129
Mo-93
Nb-93m
Nb-94
Ni-59
Ni-63
Np-237
Np-238
Np-239Pa-231
Pa-233
Pa-234m
Pb-210
Pd-107
Pu-238
Pu-239Pu-240Pu-241
Pu-242
Ra-226
Se-79
Sm-151
Sn-121m
Sn-126
Sr-90
Tc-99
Th-229Th-230Th-232
Th-234
U-233
U-234
U-235
U-236
U-237
U-238
Zr-93

92

PWRMOX – Radionuclide inventory (Bq/canister)Constr. mtrl.
Crud
UO2-matrix

1.12E+06
3.14E+08
3.67E+14
9.32E+11
9.36E+11
3.54E+12
1.99E+10
8.18E+09
2.85E+08
7.70E+11
9.82E+11
1.21E+14
6.77E+10
1.21E+10
3.47E+10
3.86E+15
8.02E+10
7.04E+12
7.50E+09
2.27E+09
7.76E+07
1.06E+11
1.13E+07
1.55E+08
1.64E+10
2.84E+10
4.21E+09
3.54E+12
2.01E+06
2.84E+10
1.92E+10
5.95E+04
1.06E+10
2.41E+14
2.56E+13
4.86E+13
2.19E+15
2.61E+11
2.11E+05
5.68E+09
2.49E+13
7.13E+11
4.50E+10
2.49E+15
1.06E+12
1.94E+04
2.89E+07
3.17E+01
1.92E+10
5.38E+06
1.02E+11
9.69E+08
1.88E+10
5.24E+10
1.92E+10
1.36E+11

2.15E+08

1.91E–01
5.57E+07
4.31E+07
1.29E+05
1.30E+05
3.90E+05

4.43E+10
0.00E+00
5.16E+06

Control rod

4.60E+13

2.49E+09
3.89E+12
7.47E+05
1.07E+05
9.86E+04
1.61E+07
8.63E+03
2.22E+03

6.52E+08
0.00E+00
1.17E+10
0.00E+00
9.52E+08
1.34E+14
5.75E+11
1.66E+11
1.72E+13

0.00E+00

1.82E+08

3.13E+06
0.00E+00
1.20E+11
6.10E+08
1.43E+08

1.69E+06

1.68E+10

0.00E+00
0.00E+00
1.39E+09
0.00E+00
6.70E+04
9.75E+07
4.10E+05
3.14E+08
2.95E+10
4.55E+03
5.84E+02
3.90E+05
3.17E–01
4.55E+03
2.95E+03
1.65E–02
3.51E+07
3.42E+06
5.29E+06
2.19E+08
2.73E+04
4.78E–02

3.35E+05

1.14E+04
4.27E–03
5.10E+00
6.57E–06
2.95E+03
9.90E–01
1.48E+04
1.65E+02
3.05E+03
5.24E+03
2.95E+03
1.30E+04

1.22E+08
4.41E+10
2.14E+08
1.63E+10
1.49E+12

0.00E+00

0.00E+00

1.17E+06
0.00E+00
3.48E+07
0.00E+00
1.93E+07

0.00E+00

1.66E+03

Total

1.12E+06
4.60E+13
3.67E+14
9.32E+11
9.36E+11
3.54E+12
6.66E+10
3.90E+12
2.91E+08
7.70E+11
9.82E+11
1.21E+14
6.77E+10
1.21E+10
3.47E+10
3.86E+15
8.02E+10
7.06E+12
7.50E+09
2.27E+09
1.15E+09
1.34E+14
5.75E+11
1.83E+11
1.88E+13
2.84E+10
4.21E+09
3.54E+12
2.01E+06
2.84E+10
1.92E+10
5.95E+04
1.06E+10
2.41E+14
2.56E+13
4.86E+13
2.19E+15
2.61E+11
2.11E+05
5.69E+09
2.49E+13
8.33E+11
4.50E+10
2.49E+15
1.06E+12
1.94E+04
2.89E+07
3.17E+01
1.92E+10
7.07E+06
1.02E+11
9.69E+08
1.88E+10
5.24E+10
1.92E+10
1.53E+11

TR-10-13

33 PWRMOX canisters:•

UO2-values from SKBdoc 1221579 (‘Nuklidlista 100320.xls’)•

Inventory for construction material from SKBdoc 1179234 by linear interpolation between 30 and
60 MWd/kg U to find inventory at 44.8 MWd/kg U.
Assumption: At any given burnup or age the construction material data is the same for PWR andPWRMOX.
(Data from the files ‘90_Ind-P30-000.xls’ and ‘91_Ind-P60-000.xls’)•

Inventory for Crud from SKBdoc 1179234 by linear interpolation between 30 and 60 MWd/kg U
to find inventory at 44.8 MWd/kg U.
Assumption: At any given burnup or age the crud data is the same for PWR and PWRMOX.
(Data from the files ‘97_Crud-P30-000.xls’ and ‘98_Crud-P60-000.xls’)•

Inventory for control rods from SKBdoc 1179234 (‘110_PWR-ss.xls’). One control rod cluster
per canister is assumed.

Table C-14. Contribution from predominant radionuclides to the total decay power at time for
encapsulation /­SKBdoc 1221579/.
Radionuclide

Decay power (W)BWR I
BWR II

BWR III

BWR MOXPWR I

PWR IIPWR IIIPWR MOX

Am- 241

270

315

211

404

294

355

238

331

Am- 243

2

3

2

3

2

4

3

3

Ba- 137m

423

381

420

349

411

349

386

387

Cm-242

1

1

1

2

Cm-243

1

1

1

1

1

1

1

1

Cm-244

80

103

147

108

87

121

185

114

1

1

Co-60

1

Cs-1342

Cs-137

127

114

126

105

123

105

116

116

Eu-154

8

4

11

5

7

3

13

10

Kr-85

3

2

4

2

3

1

4

4

Pu-238

213

278

238

232

220

316

267

216

Pu-239

21

21

16

28

21

21

16

22

Pu-240

39

45

34

54

34

42

31

41

Pu-241

2

1

2

1

2

1

2

2

Sr-90

88

74

84

70

85

66

75

78

Y-90

420

355

401

334

407

314

358

373

TR-10-13

93

Table C-15. Contribution from different isotopes in the fuel matrix to the total mass in one BWR I
and one PWR I type canister at the time for encapsulation (for isotope masses greater than orequal to 1 g/canister in either BWR I or PWR I type canister).
Isotope

BWR I
Mass(g)PWR I
Mass(g)Ag-109

188

189

Al-27

13

11

Am-241

2,358

2,569

Am-242m2

2

Am-243324

361

Ba-134

449

455

Ba-1351<1Ba-136

78

71

Ba-137

1,919

1,900

Ba-138

3,326

3,268

Br-81

53

52

C-12

17

15

Ca-40

6

5

Cd-110102

102

Cd-111

58

59

Cd-11229

28

Cd-11432

32

Cd-11612

12

Ce-140

3,210

3,143

Ce-142

2,858

2,809

Cl-35

3

3

Cm-24428

31

Cm-245

5

6

Cr-52

2

2

2,867

2,802

Cs-133
Cs-135

1,045

870

Cs-137

1,309

1,272

Dy-1607<1

Dy-1616<1

Dy-162

6

2

Dy-1638

5

Dy-164<1

2

Er-1663

4

Eu-1518

9Eu-153300

304

Eu-154

3

3

F-19

4

4

Fe-56

10

9

Gd-152

7<1
67

Gd-154132

Gd-15515

15

Gd-156

1,697

233
52

Gd-158

1,830

Gd-160

953

3

He-4

11

11

Ho-165

9

7

I-127

117

115

I-129

389

383

In-115

4

4

Kr-82

2

2

Kr-83

100

98

Kr-84

299

291

94

TR-10-13

Isotope

BWR I
Mass(g)PWR I
Mass(g)

Kr-85

6

5

Kr-86

461

451

La-139

3,109

3,056

Mg-24

2

2

Mn-55

4

3

Mo-100

2,398

2,356

Mo-95

1,914

1,880

Mo-96

117

113

Mo-97

2,062

2,023

Mo-98

2,103

2,063

N-14

29

26

Nd-14262

59

Nd-143

1,926

1,912

Nd-144

3,474

3,390

Nd-145

1,684

1,647

Nd-146

1,818

1,787

Nd-148936

922

Nd-150

456

448

Ni-58

7

6

Ni-60

3

3

1,214

1,320

Np-237
Pd-104

648

625

Pd-105

1,018

988

Pd-106945

960

Pd-107565

564

Pd-108367

368

Pd-110120121Pr-141

2,850

2,799

Pu-238375

387

Pu-23910,972

11,052

Pu-240

5,563

4,860

Pu-241

470

490

Pu-242

1,356

1,432

Rb-85

304

297

Rb-87

622

608

Rh-103

1,148

1,142

Ru-100305

311

Ru-101

1,987

1,954

Ru-102

2,051

2,017

Ru-104

1,428

1,412

Sb-121

9

9

Sb-123

12

12

Se-77

2

2

Se-78

6

6

Se-79

12

12

Se-80

32

31

Se-82

84

83

Si-28

19

17

Sm-147596

560

Sm-148

369

342

Sm-1497

9Sm-150756768Sm-15125

27

Sm-152317

304

Sm-15492

92

TR-10-1395

Isotope

BWR I
Mass(g)PWR I
Mass(g)

Sn-116

7

8

Sn-117

11

11

Sn-1189

9

Sn-11999Sn-12099Sn-12212

12

Sn-12420

20

Sn-126

47

47

Sr-86

1

<1

Sr-88

858

839

Sr-90

538

520

Tb-159

57

6

Tc-99

1,982

1,935
27

Te-12527

Te-1262

1

Te-128

221

218

Te-130

958

942

Ti-46

2

<1

Ti-47

2

<1

Ti-48

15

13

U-234

443

552

U-235

15,084

15,116

U-236

10,397

10,184

U-238

1,963,200

1,722,000

Xe-1289

9

Xe-13019

19

Xe-131

1,029

992

Xe-132

2,878

2,856

Xe-134

3,904

3,840

Xe-136

5,815

5,884

Y-89

1,146

1,121

Zn-64

25

22

Zn-66

15

13

Zn-67

2

2

Zn-68

10

9

Zr-90

878

864

Zr-91

1,506

1,474

Zr-92

1,638

1,604

Zr-93

1,790

1,754

Zr-94

1,958

1,919

Zr-96

2,051

2,011

96

TR-10-13

Appendix D
Glossary of abbreviations and branch terms
BWR

Boiling water reactor

Clab

Central interim storage facility

Clink

Central interim storage facility and encapsulation plant

Crud

Deposits on the surface of the fuel cladding

Crudact

Computer program for calculation of radionuclide inventory in crud

FGR

Fission gas release

HM

Heavy metal

IndactComputer program for calculation of induced (neutron activation)
radionuclide inventory

Keff

Effective multiplication factor, shall be less than 0.95 to avoid criticality

MOX

Mixed oxide

Origen-S

Computer program for calculation of radionuclide inventory inirradiated fuel

PWR

Pressurised water reactor

transfer canister

Canister used to transfer spent fuel assemblies between handling pool in
the interim storage part and encapsulation part of Clink

storage canister

Canister used for interim storage of spent fuel assemblies

UOX

Uranium oxide

TR-10-13

97