potential production of molybdenum-99 · 2014. 10. 1. · page | ii abstract in this study, the...

DEVELOPMENT OF A 37-ELEMENT FUEL BUNDLE FOR THE PRODUCTION OF

MOLYBDENUM-99 IN CANDU POWER REACTORS

By

Jawad Haroon

A Thesis Submitted in Partial Fulfillment

of the Requirements for the Degree of Master of Applied Science

In

The Faculty of Energy Systems and Nuclear Science

Nuclear Engineering

University of Ontario Institute of Technology

August 2014

© Jawad Haroon, 2014

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Abstract

In this study, the potential use of CANDU power reactors for the production of

Mo-99 is assessed. Five different modifications of a 37-element fuel bundle that could

be used for the production of Mo-99 in existing CANDU type power reactors are

explored. Since Mo-99 is generated by the fission of U-235 with a fission yield of 6.1%,

the proposed designs use enriched uranium in the Mo-99 producing fuel pins. The

challenge lies in designing a fuel that is able to produce significant quantities of Mo-99

while having similar neutronics and thermalhydraulic characteristics to the standard

CANDU fuel.

The proposed designs, when irradiated in the peak power channel of a CANDU

core, are shown to produce significant quantities of Mo-99 while maintaining the

necessary reactivity and power rating limits. The total Mo-99 production activities are

found to be approximately 2335, 2297, 2336, 4131 and 4268 six-day Curies per bundle

for Designs 1 to 5, respectively. The yield corresponds to approximately 19% (for

Designs 1, 2 and 3) and 34% (for Designs 4 and 5) of the world weekly demand for

Mo-99. A production cycle of 6 bundles per week (for Designs 1, 2 and 3), or 3 bundles

per week (for Designs 4 and 5) can meet the global demand of Mo-99 medical isotopes.

Keywords: Medical Isotopes, CANDU, DRAGON Code, Molybdenum-99, Six-Day Curies,

Infinite Multiplication Factor, Linear Heat Rating, Fuel Centreline Temperature

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Acknowledgement

I am grateful to my research supervisor Dr. Eleodor Nichita for giving me this

opportunity to research under his guidance. I sincerely thank him for his valuable advice

and confidence in my abilities throughout my Masters. I thank him for all the guidance

and information he provided that allowed me to start off this project, and helping me

every time I felt lost.

A special thank you to Wargha Peiman, who is a PhD candidate at the University

of Ontario Institute of Technology for his help with the DRAGON code and the many

interesting discussions we had on deciphering code manuals.

I would like to dedicate this thesis to my mother and father, Sarwat Siddiqa and

Haroon-ur-Rashid, who from an early age encouraged me to pursue my interest in

science and engineering. Thank you for all the encouragement along the way and for the

value placed on family. Thank you to my wonderful sister, Saba Haroon, who was there

for me throughout the writing process to help me proofread and give suggestions. I

would like to thank all my family and friends for their love and support throughout my

Masters, in particular, Brooke Smyth for many years of happiness and continued love,

support and inspiration.

It has been an honour and a pleasure, thank you all.

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Table of Contents

ABSTRACT ........................................................................................................................................................ II

ACKNOWLEDGEMENT ................................................................................................................................ III

TABLE OF CONTENTS .................................................................................................................................. IV

LIST OF ACRONYMS .................................................................................................................................... XII

CHAPTER 1 INTRODUCTION ...................................................................................................................... 1

1.1 INTRODUCTION TO NUCLEAR MEDICINE .................................................................................................... 3

1.1.1 WHAT IS NUCLEAR MEDICINE? .............................................................................................................................. 3

1.1.2 MEDICAL ISOTOPES AND DISEASES ....................................................................................................................... 4

1.1.3 MOLYBDENUM-99/TECHNETIUM-99M: THE MOST COMMON RADIOISOTOPE ........................................... 5

1.2 DEMAND FOR MOLYBDENUM-99 ............................................................................................................... 5

1.3 THE BIG FIVE MEDICAL ISOTOPE PRODUCING REACTORS ....................................................................... 7

1.4 INTRODUCTION TO RADIOISOTOPE PRODUCTION IN NUCLEAR REACTORS ......................................... 11

1.5 ALTERNATIVE METHODS FOR PRODUCING MOLYBDENUM-99 ............................................................ 12

1.5.1 PHOTO-FISSION PROCESS 238U(Γ,F)99MO ......................................................................................................... 12

1.5.2 NEUTRON ACTIVATION OF MOLYBDENUM-98 ................................................................................................. 12

1.5.3 NEUTRON EMISSION FROM MOLYBDENUM-100 ............................................................................................. 13

1.6 RADIOACTIVE DECAY OF MOLYBDENUM-99/TECHNETIUM-99M ....................................................... 14

1.7 CONSIDERATIONS IN THE PRODUCTION OF MOLYBDENUM-99 IN NUCLEAR REACTORS .................... 16

1.7.1 MOLYBDENUM-99 TARGET PROCESSING TECHNIQUES ................................................................................. 18

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1.7.2 TECHNETIUM-99M GENERATORS....................................................................................................................... 21

1.7.3 SPENT FUEL CONSIDERATIONS ........................................................................................................................... 23

1.8 CANDU REACTORS AND THE DRAGON LATTICE CODE ...................................................................... 26

1.8.1 USING CANDU REACTORS FOR MOLYBDENUM-99 PRODUCTION ............................................................... 26

1.8.2 THE DRAGON CODE ............................................................................................................................................ 30

CHAPTER 2 PRODUCTION OF MOLYBDENUM-99 IN NUCLEAR REACTORS: PRESENT AND

FUTURE ........................................................................................................................................................... 33

2.1 CURRENT MOLYBDENUM-99 SUPPLY CHALLENGES .............................................................................. 34

2.2 CONVERSION OF MOLYBDENUM-99 TARGETS FROM HIGHLY-ENRICHED URANIUM TO LOW-

ENRICHED URANIUM ....................................................................................................................................... 35

2.3 NEW PRODUCERS OF MOLYBDENUM-99 WORLDWIDE ........................................................................ 37

2.4 MOLYBDENUM-99 PRODUCTION IN POWER REACTORS ....................................................................... 38

CHAPTER 3 METHODOLOGY FOR THE DEVELOPMENT OF A MOLYBDENUM-99

PRODUCING 37-ELEMENT FUEL BUNDLE DESIGN ........................................................................... 41

3.1 DESIGN OBJECTIVES .................................................................................................................................. 41

3.2 DESIGN PLAN AND STRATEGY .................................................................................................................. 42

3.2.1 SATISFYING DESIGN REQUIREMENTS ................................................................................................................. 45

3.3 REACTOR OPERATING CONDITIONS AND ASSUMPTIONS ....................................................................... 51

3.3.1 REACTOR FUELLING CONSIDERATIONS ............................................................................................................. 52

3.4 DESIGN OPTIONS AND METHODOLOGY ................................................................................................... 54

3.4.1 METHOD FOR CALCULATING THE LINEAR HEAT RATING ............................................................................... 57

3.4.2 METHOD FOR CALCULATING THE RADIAL TEMPERATURE PROFILE ............................................................ 60

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3.4.3 METHOD FOR CALCULATING THE MASS AND ACTIVITY OF MOLYBDENUM-99 PRODUCED ..................... 69

CHAPTER 4 ASSESSMENT RESULTS AND DISCUSSION ................................................................... 73

4.1 TASK 1 - COMPARISON OF INFINITE MULTIPLICATIVE CONSTANTS ..................................................... 78

4.2 TASK 2 - COMPARISON OF LINEAR HEAT RATING ................................................................................. 85

4.3 TASK 3 - COMPARISON OF RADIAL TEMPERATURE PROFILES .............................................................. 92

4.4 MASS AND ACTIVITIES OF MOLYBDENUM-99 ...................................................................................... 113

CHAPTER 5 CONCLUDING REMARKS AND FUTURE WORK ........................................................ 117

REFERENCES ............................................................................................................................................... 120

APPENDICES ............................................................................................................................................... 135

APPENDIX A – POTENTIAL PRODUCERS OF REACTOR PRODUCED MOLYBDENUM-99............................ 135

A.1 MULTIPURPOSE APPLIED PHYSICS LATTICE EXPERIMENT (MAPLE) .......................................................... 135

A.2 AQUEOUS HOMOGENEOUS REACTOR (AHR) ..................................................................................................... 137

A.3 OPEN POOL AUSTRALIAN LIGHT-WATER (OPAL) REACTOR ........................................................................ 138

A.4 JULES HOROWITZ REACTOR (RJH) ...................................................................................................................... 140

A.5 CHINA ADVANCED RESEARCH REACTOR (CARR) ............................................................................................ 142

A.6 UNIVERSITY OF MISSOURI RESEARCH REACTOR CENTER (MURR) .............................................................. 143

A.7 SANDIA NATIONAL LABORATORIES (SNL) ........................................................................................................ 144

A.8 OTHER POTENTIAL SUPPLIERS OF REACTOR PRODUCED MOLYBDENUM-99 .............................................. 148

A.9 NEW APPROACHES IN DIAGNOSTIC NUCLEAR MEDICINE ................................................................................ 150

APPENDIX B – MATERIALS, LATTICE SPECIFICATIONS AND IMPORTANT PARAMETERS FOR 37-ELEMENT

CANDU FUEL BUNDLE CALCULATIONS ....................................................................................................... 151

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APPENDIX C – INITIAL ESTIMATES OF ENRICHED FUEL REGION THICKNESS AND PIN RADIUS FOR UO2

AND U-METAL FUEL DESIGNS BASED ON URANIUM MASS BALANCING .................................................... 153

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List of Figures

Figure 1-1: Approximate Global Demand for Molybdenum-99 ............................................................................................ 7

Figure 1-2: Principal Decay Scheme of Molybdenum-99 ...................................................................................................... 16

Figure 1-3: Buildup of Molybdenum-99 Radioactivity for a Neutron Flux Typical of a CANDU-6 ....................... 18

Figure 1-4: Global Supply Chain of Molybdenum-99 and Utilization Schematics ....................................................... 20

Figure 1-5: Decay of Molybdenum-99 and Buildup of Technetium-99m Activity after Each Elution,

Calculated for 7 Days .......................................................................................................................................................................... 22

Figure 1-6: Generic Flow Chart of Waste Streams Generated from Molybdenum-99 Production ........................ 25

Figure 1-7: CANDU Reactor Face ................................................................................................................................................... 28

Figure 1-8: A Typical CANDU Fuel Channel................................................................................................................................ 29

Figure 3-1: A Pictorial Representation of the Proposed Designs Assessed for Molybdenum-99 Production .... 44

Figure 3-2 37-Element CANDU Fuel Bundle and the Infinite Lattice Cell Representation of Design 2.............. 57

Figure 3-3: A Plot of UO2 Thermal Conductivity Showing Variation Due to Temperature ..................................... 65

Figure 4-1: Radial Pin Power Profile for Reference CANDU 37-Element Bundle ........................................................ 74

Figure 4-2: Reference CANDU 37-Element Bundle Radial Temperature Profile of Ring 4 Pin Using IAEA and

Constant Thermal Conductivity for UO2 ...................................................................................................................................... 77

Figure 4-3: Design 1 Unitcell Lattice k-infinity Versus Variations in Radius of Enriched Central Target

Region (Fuel Pin Radius 6.075 mm) .............................................................................................................................................. 80

Figure 4-4: Design 2 Unitcell Lattice k-infinity Versus Variations in Thickness of Enriched Annular Target








Figure 4-8: Design 1 Bundle Radial Pin Power Profile ........................................................................................................... 87

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Figure 4-9: Design 2 Bundle Radial Pin Power Profile ........................................................................................................... 88

Figure 4-10: Design 3 Bundle Radial Pin Power Profile ........................................................................................................ 89



Figure 4-13: Design 1 Radial Temperature Profile of Ring 4 Pin Using IAEA Thermal Conductivity for UO2 . 97

Figure 4-14: Design 1 Radial Temperature Profile of Ring 4 Pin Using a Constant Thermal Conductivity for

UO2 (0.024 W/cm∙K) ............................................................................................................................................................................ 98

Figure 4-15: Design 2 Radial Temperature Profile of Ring 4 Pin Using IAEA Thermal Conductivity for

UO2 ........................................................................................................................................................................................................... 101


UO2 (0.024 W/cm∙K) ......................................................................................................................................................................... 102


U-Metal (0.43 W/cm∙K) ................................................................................................................................................................... 105

Figure 4-18: Design 4 Radial Temperature Profile of Ring 4 Pin Using IAEA Thermal Conductivity for

UO2 ........................................................................................................................................................................................................... 108


UO2 (0.024 W/cm∙K) ......................................................................................................................................................................... 109


U-Metal (0.43 W/cm∙K) ................................................................................................................................................................... 112

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List of Tables

Table 1-1: Worldwide Production Capacity of Molybdenum-99 ([11], [12], [13], [14]) ............................................. 9

Table 1-2: Small-Scale Producers of Molybdenum-99 in 2011 ([8], [11], [14]) ........................................................... 10

Table 1-3: Status of CANDU reactors and CANDU-Derivatives in 2014 ([28], [29]) .................................................. 27

Table 4-1: Reference CANDU 37-Element Bundle Data Structures for Radial Pin Power Profile

Calculations............................................................................................................................................................................................. 74

Table 4-2: Reference CANDU 37-Element Bundle Normalized Heat Density and Linear Power Rating in Ring

4 Sub-regions .......................................................................................................................................................................................... 75

Table 4-3: Reference CANDU 37-Element Bundle Numerical Integration Results Using IAEA and Constant

Thermal Conductivity for UO2 .......................................................................................................................................................... 76

Table 4-4: Design 1 Unitcell Lattice k-infinity for Variations in Radius of Enriched Central Region .................. 80

Table 4-5: Design 2 Unitcell Lattice k-infinity for Variation in Thickness of Enriched Annular Region ............ 81




Table 4-9: Design 1 Data for Radial Pin Power Profile Calculations ............................................................................... 87

Table 4-10: Design 2 Data for Radial Pin Power Profile Calculations ............................................................................. 88




Table 4-14: Thermophysical Properties of Coolant Using NIST REFPROP [46] ........................................................... 94

Table 4-15: CANDU Specific Important Heat Transfer Parameters [46] ........................................................................ 94

Table 4-16: Design 1 Normalized Heat Density and Linear Power Rating in Ring 4 Sub-regions ....................... 95

Table 4-17: Design 1 Numerical Integration Results Using IAEA and Constant Thermal Conductivity for

UO2 .............................................................................................................................................................................................................. 96

Table 4-18: Design 2 Normalized Heat Density and Linear Power Rating in Ring 4 Sub-regions ....................... 99

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UO2 ........................................................................................................................................................................................................... 100

Table 4-20: Design 3 Normalized Heat Density and Linear Power Rating in Ring 4 Sub-regions .................... 103

Table 4-21: Design 3 Numerical Integration Results Using Constant Thermal Conductivity for U-Metal ..... 104



UO2 ........................................................................................................................................................................................................... 107


Table 4-25: Design 5 Numerical Integration Results Using Constant Thermal Conductivity for U-Metal ..... 111

Table 4-26: Design 1 Molybdenum-99 Production Results ............................................................................................... 114





Table A-1: Potential Future Suppliers of Reactor Produced Molybdenum-99 [13] ................................................. 149

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List of Acronyms

ACRR Annular Core Research Reactor

AECL Atomic Energy of Canada Limited

AHR Aqueous Homogeneous Reactor

ANSTO Australian Nuclear Science and Technology Organization

B&W Babcock & Wilcox

CANDU CANada Deuterium Uranium

CARR China Advanced Research Reactor

CEA Atomic Energy Commission

CHF Critical Heat Flux

CNSC Canadian Nuclear Safety Commission

DOE Department of Energy

FDA Food and Drug Administration

HEU Highly-Enriched Uranium

HFR High Flux Reactor

IAEA International Atomic Energy Agency

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LEU Low-Enriched Uranium

MNR McMaster Nuclear Reactor

MURR University of Missouri Research Reactor

NRU National Research Universal

NU Natural Uranium

OPAL Open Pool Australian Light-Water Reactor

RJH Jules Horowitz Reactor

SNL Sandia National Laboratories

SNM Society of Nuclear Medicine

US-NRC Unites States Nuclear Regulatory Commission

WLUP WIMS-D Library Update

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Chapter 1

Introduction

In nuclear medicine, radioactive nuclides are used to obtain functional

information about a patient’s organs and to diagnose and treat various medical

conditions [1]. The most common radioisotope used in nuclear medicine is

technetium-99 (Tc-99m), which is produced through the radioactive decay of

molybdenum-99 (Mo-99). Tc-99m is used in approximately 30 million procedures per

year, accounting for 80% of all nuclear medicine procedures worldwide [2].

Mo-99 is most commonly and effectively generated in nuclear reactors by the

fission reaction of U-235 with a yield of 6.1% [3]. Mo-99 production in fission reactors

is usually achieved by irradiation of specially designed highly-enriched uranium (HEU)

targets located inside irradiation rigs positioned at special high-flux sites in the core.

These HEU targets typically have a U-235 enrichment of 90 wt% or higher. Although

some Mo-99 is produced in natural uranium (NU), the NU fuel composition is not suitable

for producing significant quantities of Mo-99 since the ratio of Mo-99 yield to fuel volume

processed is generally low. Therefore, the use of enriched fuel is desirable from the

perspective of Mo-99 production.

When uranium is enriched to less than 20% by weight of U-235, it is considered

low-enriched enriched uranium (LEU). Concerns over weapon proliferation and nuclear

security have led to advanced discussions between experts of the field with the aim of

using LEU instead of HEU as reactor fuel and target material.

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As of 2014, the National Research Universal (NRU) reactor in Chalk River,

Ontario, produces approximately 40% of the global demand for Mo-99 [4]. The NRU is

57 years old and is scheduled to cease Mo-99 production in the year 2016. Consequently,

alternative production methods and radioisotope production facilities need to be

developed. One possibility is to irradiate specially designed fuel in the operating

CANDU®1 reactors; this is the main driver of the work presented in this thesis.

A CANDU reactor with its pressure channel design and its ability to fuel online

has the potential to produce large quantities of Mo-99, while maintaining its power

production targets. The production and extraction of Mo-99 is not possible in most other

power plants due to batch fuelling. The online refuelling capabilities in CANDU, with its

flexible and compact fuel design give CANDU reactors an edge over pressure-vessel light

water reactors which are generally refuelled in batches every 18 to 24 months.

The goal of this thesis is to evaluate various modified 37-element fuel bundles

specifically made for CANDU reactors with the intent to generate a significant portion of

the world’s demand for Mo-99 without significantly changing the physics or fuel bundle

safety characteristics. Several variations of the 37-element fuel bundle are analyzed

using the DRAGON lattice code. Strict design acceptance criteria are used, consisting of

close proximity of the fresh fuel k-infinity value to that of the reference CANDU bundle

1 CANDU is a trademark of Atomic Energy of Canada Limited used under licence by Candu Energy

Inc.

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(Δρ < 1 mk) and maintaining existing margins to linear heat rating limits. An assessment

of the fuel centreline temperature is also performed for the modified fuel bundle pins.

The strict design constraints allow the station to maintain existing operating

practices and eliminate the need for supplemental fuel safety analysis or full core

neutronics calculations.

1.1 Introduction to Nuclear Medicine

1.1.1 What is Nuclear Medicine?

Nuclear medicine uses small quantities of unstable nuclei administered to a

patient to study the function of specific organs in the body or to treat tumours and other

diseases ([1], [4]). The field of nuclear medicine can be divided into two branches:

diagnostics and therapeutics.

Diagnostics: A radiopharmaceutical consisting of a radioisotope bound to a

substrate is administered to a patient. The substrate is what governs the distribution of

the radiopharmaceutical in the body and is chosen such that it is predominantly

accumulated in the organ that needs to be investigated. The amount of accumulation in

the specific organ gives information about the health and function of the organ. The

gamma-emitting radioactive isotopes is what allows the distribution of the

radiopharmaceutical to be visualized using an external detector which produces a

diagnostic image. Images are then analyzed to detect partial or excessive absorption of

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the isotope as an indication of organ malfunction [4]. Nuclear medicine imaging

provides a view of the position and concentration of the radioisotope which depends

primarily on an organ’s functioning rather than on its anatomy.

Therapeutics: Radioisotopes are injected into a patient in higher doses. The

isotopes emit very energetic photons or particles which can destroy targeted cells such

as cancer cells.

1.1.2 Medical Isotopes and Diseases

Medical isotopes are used routinely in the diagnosis and treatment of various

diseases at hospitals and radiation clinics. The following list includes, but is not limited

to some of the health conditions where various medical isotopes are used for monitoring

and evaluation [2]:

Heart disease

Level of functioning of brain, heart, lungs, kidneys and other organs

Tumours

Progression of cancer – spreading to bones

Hormonal disorders, ex. thyroid disease

Over hundreds of radiopharmaceutical products are used at hospitals and

radiation clinics in North America every day. About 95% of the radiopharmaceuticals

are used for medical diagnosis, while the rest are used for therapeutic purposes ([5], [6]).

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1.1.3 Molybdenum-99/Technetium-99m: The Most Common Radioisotope

The most commonly used radioisotope in diagnostic nuclear medicine is Tc-99m,

which is produced directly from the radioactive decay of Mo-99 ([5], [7]). Tc-99m is used

in approximately 30 million procedures per year, accounting for 80% of all nuclear

medicine procedures worldwide [2].

Mo-99 and Tc-99m do not exist naturally. Mo-99 is most commonly and

effectively generated by the fission of U-235 in nuclear reactors with a fission yield of

6.1% [3]. The reaction shown in Eq. (1.1) corresponds to the production of Mo-99 from

the fission of U-235, which occurs for approximately 6.1% of all fission reactions.

nSnMonU 4133

50

99

42

235

92 (1.1)

Mo-99 is formed from the nuclear fission of U-235 and consequently exists as a

fission by-product in uranium targets. These targets are specifically irradiated for

radioisotope production in nuclear reactors. The Mo-99 is then extracted in dedicated

processing facilities and placed in generators that are sent to various medical

institutions around the globe.

1.2 Demand for Molybdenum-99

Every week, approximately 300 therapeutic doses of medical isotopes and 30000

diagnostic treatments are performed in Canadian hospitals [2]. Over 10000 hospitals

worldwide use radioisotopes of which 90% are used strictly for diagnosis [4].

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In developed countries, which account for 26% of the world population, the

relative use of diagnostic nuclear medicine among all health procedures is 1.9% per year,

one tenth of which involves therapy using radiopharmaceuticals [4]. In the United

States, there are approximately 18 million nuclear medicine procedures per year

(population of 311 million), and there are about 10 million procedures performed in

Europe (population of 500 million). In Australia, there are approximately 560000

procedures performed per year (population of 21 million) out of which 470000 of these

procedures are done using reactor produced isotopes [4].

In North America, the demand for Mo-99 ranges from 5000 to 7000 six-day

Curies per week with an estimated annual growth rate between 3 and 5% [8]. Canada

accounts for approximately 1/11 of the total North American consumption. A

breakdown of the global demand of Mo-99 in 2009 is shown in Figure 1-1.

Curie is the unit of radioactivity (1 Curie = 3.7x1010 disintegrations per second or

Becquerels, Bq). A six-day Curie is a unit of measure most commonly used in the industry

which equates to the amount of Mo-99 remaining in a Tc-99m generator six days after

shipment from the producer’s facility. In other words, the producers usually calibrate a

shipment value to the activity of the Mo-99 isotopes 6 days after it leaves the production

facility [8].

The worldwide demand for Mo-99 has reached approximately 12000 six-day

Curies per week [7]. When accounting for the processing time and efficiency, 12000 six-

day Curies is equivalent to about 77000 end-of-bombardment Curies which is 160 mg of

Mo-99 at the end of the irradiation stage [8]. The demand for medical isotopes for

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diagnosis and therapy purposes is expected to grow over 10% per year worldwide ([4],

[9]).

Figure 1-1: Approximate Global Demand for Molybdenum-992

1.3 The Big Five Medical Isotope Producing Reactors

Currently, five nuclear reactors are producing approximately 95% of the world’s

supply of Mo-99 [11]. The five are the Chalk River National Research Universal (NRU)

reactor in Canada, the High Flux Reactor (HFR) in the Netherlands, BR2 in Belgium,

OSIRIS in France, and the SAFARI-1 in South Africa. In addition to the production of

2 Data taken from Reference [10]

44%

22%

14%

4%

16%

United States Europe Japan Canada Other

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medical isotopes, these reactors also provide multiple services such as nuclear fuel

testing and material research.

These big five industrial scale research reactors producing most of the world’s

Mo-99 supply are government-owned and funded. The global market share of these five

major isotope producing reactors is shown in Table 1-1. The Canadian NRU reactor and

the HFR in Netherlands, each produce approximately 30% to 40% of the global demand

for Mo-99. The other two European reactors and the reactor in South Africa combined,

produce approximately 36% of the Mo-99 supply. The smaller reactors around the

world support approximately 4% of the world’s demand of Mo-99.

As of 2014, Canada’s NRU in Chalk River is 57 years old, HFR in Petten,

Netherlands is 53 years old, SAFARI-1 in Pelindaba, South Africa is 49 years old, BR-2 in

Mol, Belgium is 53 years old, and OSIRIS in Saclay, France is 48 years of age [12].

Table 1-1 also provides additional information on these five reactors (e.g., thermal

powers and neutron fluxes, and estimated reactor stop date).

The majority of the smaller Mo-99 producing reactors around the globe

accommodate local or regional medical isotope markets. Some regional producers of

Mo-99 are Australia, Argentina, Germany, Indonesia, and Poland as shown in Table 1-2.

The research reactors in the aforementioned countries do not produce the commercial

quantities necessary to compensate for the closure of one or more of the big five reactors.

However, they can act as a short term make-up when one of the big five reactors

experiences an outage. The regional reactors are listed in Table 1-2.

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Table 1-1: Worldwide Production Capacity of Molybdenum-99 ([11], [12],

[13], [14])

Reactor Approximate Market Share (%)

Thermal Power (MW)

Thermal Neutron Flux (n·cm-2·s-1)

Target Type

Estimated Stop Date

National Research Universal (NRU) (Canada)

30% to 40% 135 4x1014 HEU 2016

High Flux Reactor (HFR) (Netherlands)

30% to 40% 45 2.7x1014 HEU 2022

SAFARI-1 (South Africa)

15% 20 2.4x1014 HEU 2025

BR2 (Belgium) 14% 100 1x1015 HEU 2026 OSIRIS (France) 7% 70 1.7x1014 HEU 2018 Rest of the world

4% - - -

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Table 1-2: Small-Scale Producers of Molybdenum-99 in 2011 ([8], [11], [14])

Reactor Typical Capacity (Six-day Curies)

Thermal Power (MW)

Thermal Neutron Flux (n·cm-2·s-1)

Target Type

OPAL (Australia) 20 3x1014 LEU

RA-3 (Argentina) 200 5 4.8x1013 LEU MARIA (Poland) 30 3.5x1014 HEU LWR-15 (Czech Republic)

10 1.5x1014 HEU

WWR-TS (Russia) 15 1.8x1014 HEU IRT-T (Russia) 6 1.4x1014 HEU FRM-II (Germany) 20 8x1014 HEU ETTR-2 (Egypt) 250 22 2.8x1014 LEU RSG-GAS (Indonesia)

150 30 2.5x1014 HEU

PARR-1 (Pakistan)

20 10 1.7x1014 HEU

RECH-1 (Chile) 250 5 7x1013 LEU

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1.4 Introduction to Radioisotope Production in Nuclear Reactors

Neutrons in nuclear reactors drive the nuclear reactions that must take place for

the successful production of various radioisotopes. The neutron flux typically achieved

in nuclear reactors is several orders of magnitude higher when compared to neutron

generators and other isotopic neutron sources [7], for example ~1014 n·cm-2·s-1 for a

typical nuclear reactor versus 1010 n·cm-2·s-1 for a typical neutron source. The quantity

and activity of radioisotopes is directly proportional to the neutron flux. This explains

why nuclear reactors that have high neutron flux values hold a strategic advantage in the

production of medical radioisotopes.

Although all reactors are neutron sources; not all are suitable for Mo-99

production. The factors that determine whether a reactor is suitable for Mo-99

production are:

(i) neutron flux available for target irradiation,

(ii) volume of the irradiation rigs into which targets can be placed, and

(iii) in the case where targets are fissionable isotopes of enriched uranium, a further

consideration is heat removal from the targets [15].

Currently, nuclear reactors are producing Mo-99 and its decay product Tc-99m

through irradiation of HEU targets. After spending several days in the reactors, these

irradiated targets are sent for processing to dedicated processing facilities [7].

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1.5 Alternative Methods for Producing Molybdenum-99

A nuclear reactor is not a requirement when it comes to producing Mo-99. An

accelerator can also be utilized to generate Mo-99 by photo-fission of the U-238 atoms

present in NU. Alternative options for the production of Mo-99 include eliminating

uranium from the process by using other isotopes of molybdenum and converting them

to Mo-99. This consists of four possible reactions: 98Mo(n,γ)99Mo (neutron activation)

occurring inside a reactor, 100Mo(γ,n)99Mo, 100Mo(p,pn)99Mo and 100Mo(p,2n)99mTc

(neutron emission) using accelerators.

1.5.1 Photo-fission Process 238U(γ,f)99Mo

In this method, a high intensity beam of photons is generated and directed at

U-238 [16]. This reaction results in nearly the same 6.1% fission yield for Mo-99 as the

fission yield of U-235 in nuclear reactors. The cross-section for this reaction is about

1000 times smaller than the neutron fission cross-section of U-235. As a result, a

relatively large number of accelerators would be required to produce significant

amounts of Mo-99. One estimate is that half a dozen accelerators are needed to supply

30% to 50% of North American demand [8].

1.5.2 Neutron Activation of Molybdenum-98

The reaction, 98Mo(n,γ)99Mo (neutron activation), requires thermal (~ 0.025 eV)

or epithermal (0.025 – 1.0 eV) neutrons. However, only a small portion of this target is

converted to Mo-99 because the cross-section of this reaction is small (σthermal ~ 0.14

barns). In addition, although neutron activation of molybdenum generates the desired

Page | 13

radionuclide with little to no waste stream, Mo-99 produced through this method is not

“carrier-free” as it contains Mo-98 (natural molybdenum contains approximately 24%

of Mo-98) which is chemically identical to Mo-99 and behaves as a contaminant.

The Mo-99 produced through neutron activation has low specific activity and

requires a large number of targets to produce significant quantities of Mo-99. A nuclear

reactor would still be required to provide the necessary neutron flux for this method.

Furthermore, the potential exists for increased elution of undesirable Mo-98 when

formulating patient dosages [8].

1.5.3 Neutron Emission from Molybdenum-100

The three neutron emission methods (100Mo(p,pn)99Mo, 100Mo(γ,n)99Mo and

100Mo(p,2n)99mTc) require accelerators that can generate very high intensity beams of

protons (500 μA) or electrons, in order to create photons powerful enough to overcome

the significantly smaller cross-sections for these reactions.

When the 100Mo(γ,n)99Mo reaction is used, high energy photons known as

Bremsstrahlung radiation are produced by the electron beam as it interacts and loses

energy in the ‘converter’ target. The photons are subsequently used to irradiate another

target material (Mo-100) placed just behind the converter to produce Mo-99 via neutron

emission.

The 100Mo(p,2n)99mTc neutron emission method, results in the direct production

of Tc-99m which has a significant disadvantage due to its short half-life. It is not an ideal

method to produce Tc-99m because any need to ship the Tc-99m generators over

Page | 14

distances would reduce the usefulness to the end-user and by extension, to the patient

[8].

1.6 Radioactive Decay of Molybdenum-99/Technetium-99m

Mo-99 decays into Tc-99m, which is then extracted from the generators and

combined with other reagents to form a radiopharmaceutical that can be given to

patients. Tc-99m is an ideal radioisotope for diagnostic nuclear medicine ([7], [10], [17])

because of its physical and chemical properties. The 6.01-hour half-life of Tc-99m is

neither too long nor too short for a medical procedure. Its parent radionuclide, Mo-99,

has a 66-hour half-life which provides sufficient time for processing and long-distance

transportation, and thus permits shipping under the form of a Mo-99/Tc-99m generator

[5].

Conversely, due to its short half-life, Tc-99m cannot be stockpiled or produced

directly in nuclear reactors or through neutron activation as a final product, nor can it

be shipped to hospitals and radiation clinics around the globe. Instead, Tc-99m is

obtained through elution with a sterile saline solution from an alumina column

containing its parent isotope Mo-99. The alumina column assembly is known as a

Tc-99m ‘generator’ [3].

Another desirable property of Tc-99m, is the energy of the emitted gamma

radiation, which is in a range that allows a clear image to be produced in a gamma

camera. In addition, since 89% of Tc-99m emits pure gamma radiation that is energetic

Page | 15

enough to pass through human body without significant attenuation, this keeps the total

dose administered to patients low ([1], [18]).

Figure 1-2 shows the decay sequences of Mo-99 to Tc-99 with the respective half-

lives and yields. Mo-99 disintegrates with a half-life of 66 hours to either Tc-99m or

Tc-99 by emitting a beta particle. The “m” in Tc-99m stands for ‘meta-stable’. The

Tc-99m radionuclide is unusual since it undergoes a gamma decay into Tc-99 ground

state rather than decaying instantaneously, as is the case with most excited

radionuclides produced through β- or β+ decay [19]. As depicted in Figure 1-2, Tc-99m

decays by isomeric transition. The principal gamma-emission associated with the

transition from the meta-stable state to the ground state of Tc-99 emits a 140-keV

photon [5]. Using a scintillator that is made up of special plastic or NaI-Tl crystals, the

photon is captured by the detector, and converted into visible light photons which are

then converted by a photomultiplier tube into electrical pulses and counted

electronically [20].

Page | 16

Figure 1-2: Principal Decay Scheme of Molybdenum-993

1.7 Considerations in the Production of Molybdenum-99 in Nuclear

Reactors

The activity of Mo-99 produced in a target is a function of irradiation time, the

thermal neutron fission cross-section for U-235, the thermal neutron flux induced on the

target, the mass of U-235 in the target, and the half-life of Mo-99 [22]. For a typical

reactor with thermal neutron fluxes on the order of 1014 n·cm-2·s-1, irradiation times of

about 15 to 20 days are sufficient to achieve the maximum possible activity of Mo-99 in

the targets. Beyond this equilibrium stage, the amount of Mo-99 produced in the targets

3 Reproduced from Reference [21]

Page | 17

approximately balances the amount of Mo-99 being lost to radioactive decay [8].

Additional irradiation time is also uneconomical due to the fact that there would be

additional accumulation of other undesirable fission products, adding to the radioactive

fields from the targets, and the wastes that ultimately require disposal.

In theory, for a target irradiated by a thermal neutron flux inside a reactor, the

rate of production can be represented by the following mathematical relationship:

NNdt

dNT (1.2)

For Mo-99 production in U-235 targets, Eq. (1.2) can be solved to determine the

specific activity of radioactive atoms at time ‘t’:

)/()1(6.0 99

gBqeA

fS tfMo

(1.3)

In Eqs. (1.2) and (1.3), NT is the total number of U-235 atoms in the target, N is

the number of Mo-99 atoms produced as a fission product, 99Mof is the fraction of fission

products leading to the production of Mo-99 (6.1%), f is the microscopic fission cross-

section, is the thermal neutron flux, and is the Mo-99 decay constant. Figure 1-3

illustrates the buildup of Mo-99 activity for a target irradiated in a thermal neutron flux

of approximately 2×1014 n·cm-2·s-1, which is typical of a CANDU reactor.

Page | 18

Figure 1-3: Buildup of Molybdenum-99 Radioactivity for a Neutron Flux

Typical of a CANDU-6

1.7.1 Molybdenum-99 Target Processing Techniques

After irradiation in the reactor for several days, the HEU or LEU target is

chemically processed by one of the two processes:

i) acid dissolution and molybdenum separation process (Cintichem process), or

ii) alkaline dissolution process.

The acid dissolution was developed by Cintichem Incorporated. The technology

was later purchased by the US Department of Energy (DOE) following the

decommissioning of Cintichem’s Tuxedo, NY reactor. The facility is currently focused on

proposing modifications to convert to an LEU target for US supplies of medical isotopes.

0.00E+00

2.00E+12

4.00E+12

6.00E+12

8.00E+12

1.00E+13

1.20E+13

1.40E+13

1.60E+13

1.80E+13

2.00E+13

0 100 200 300 400 500 600 700

Spe

cifi

c A

ctiv

ity

of

Mo

-99

(B

q/g

)

Time (Hours)

Page | 19

A proprietary Nordion Incorporated acid-dissolution process is currently used

for the isotope production from NRU. The process is believed to be very similar to the

Cintichem process according to available sources [20]. A brief overview of the acid

dissolution process is provided below.

The alkaline dissolution process is generally used for targets that contain

aluminum and is currently being used by all major isotope producers except Nordion

Incorporated (formerly known as MDS Nordion).

Following several days of irradiation, the targets are removed and allowed to cool

in water for up to half a day. The targets are then transferred to an associated “hot cell”

facility. The hot cells are cabins which are shielded by lead and concrete, and are able to

withstand the decay heat and intense radiation fields of the targets. The time it takes for

the processing of the targets in the hot cells is optimized in order to minimize Mo-99

losses due to radioactive decay. Approximately 1% of the generated Mo-99 is lost to

decay per hour after end-of-bombardment inside a nuclear reactor [23]. The processing

time required following irradiation inside a reactor means that a significant amount of

Tc-99m will eventually be lost.

In the hot cell facility, the cladding is punctured and volatile fission products such

as Kr-85 and Xe-133 are removed. The target assembly is dissolved in hot nitric acid,

forming a nitrate solution containing uranium, molybdenum, and other fission by-

products. The solution is poured through an alumina (Al2O3) column that adsorbs the

nitrates. The column is then washed with additional nitric acid to elute excess uranium

and other fission by-products, leaving the molybdenum bounded within the column

Page | 20

matrix. Sodium hydroxide is then added to the column which elutes purified Mo-99 [5].

This process typically yields recovery greater than 85-90% of the Mo-99 in the targets

[8].

The removed Mo-99 is immediately shipped to Covidien in Mansfield,

Massachusetts or Lantheus Medical Imaging in North Billerica, Massachusetts, where

they manufacture Tc-99m generators. These Tc-99m generators are used to extract the

decayed daughter Tc-99m isotopes from the parent Mo-99. Finally, the generators are

then shipped to destinations around the globe as depicted in Figure 1-4.

Figure 1-4: Global Supply Chain of Molybdenum-99 and Utilization

Schematics4


Page | 21

1.7.2 Technetium-99m Generators

Manufacturing of a Tc-99m generator generally consists of adsorption of Mo-99

onto an alumina column. Positively charged beads are used in the alumina column to

adsorb the Mo-99 as Mo99O42- [17]. When the Mo-99 decays it forms pertechnetate TcO4-.

Due to its single charge, pertechnetate is less tightly bound to the alumina. The column

has a higher affinity for the Mo-99 than the Tc-99m ensuring the preferential elution of

Tc-99m when a saline solution is poured through the column. Breakthrough of small

quantities of Mo-99 still occurs. The amount of breakthrough is one of the quality control

parameters for manufacturing Tc-99m generators and is regularly monitored [17].

Due to the short half-life of Mo-99 (66 hour), the complete manufacturing process

must take place quickly. For this reason, most nuclear medicine institutions rely on

weekly shipments of Tc-99m generators to meet their demands. The supply chain is

currently capable of supplying Tc-99m from reactor extraction to the patient in less than

48 hours in an ideal case.

When the supply reaches the hospital or the radiation clinic, a radiopharmacist

calculates the amount of Tc-99m generated by disintegration of Mo-99 nuclides based

on the transient equilibrium that exists between the two nuclides. For commercial kits

such as Cardiolite®, the Tc-99m is eluted from the column with sterile saline (0.9%

NaCl), as sodium pertechnetate (Na99mTcO4), and is then mixed with other reagents into

a final form that can be administered to patients. The column is re-used, or “milked”,

until most of the Mo-99 has decayed and the column is exhausted [5]. Figure 1-5 shows

the “milking” process of Tc-99m from its first elution.

Page | 22

TechneLite® is a Tc-99m generator that is produced by Lantheus Medical Imaging

Incorporated. Dosage from TechneLite can range from 0.08 mCi/kg for pediatric thyroid

imaging to 0.28 mCi/kg for blood pool imaging of the heart [24]. Cardiolite® is another

Tc-99m generator produced by Lantheus Medical Imaging that is widely used in cardiac

stress test applications.

Figure 1-5: Decay of Molybdenum-99 and Buildup of Technetium-99m Activity

after Each Elution, Calculated for 7 Days5


Page | 23

1.7.3 Spent Fuel Considerations

It is important to consider various concerns that generally arise in the production

of Mo-99 from the perspective of waste management. Management of waste starts from

the planning, designing and construction phase of the Mo-99 production facility and

continues through the operational phase. Waste management strategy has to also

account for the route taken by the spent targets and long-lived fission products.

Establishment and implementation of an effective and safe waste management strategy

is generally a licensing requirement which is approved and regulated by the authorities.

Target fabrication, irradiation, and extraction of Mo-99 from targets generate

significant quantities of chemical and radioactive waste. Waste is generated as solids,

liquids, and gases, and can include material in the low, intermediate and highly

radioactive waste categories [25]. Initial treatment of waste streams is usually done at

the production site, from where it is transferred to short or long term storage facilities.

Treatment of gaseous waste is generally carried out in the production facility on site

while treatment of solid and liquid radioactive waste can be performed off-site [25].

Figure 1-6 is a generic flow chart of waste streams generated from Mo-99 production.

The conversion from HEU to LEU targets, which is underway at most of the

current Mo-99 producing facilities, would generate even larger quantities of waste. If

current target dissolution processes are maintained, the existing waste storage facilities

may not be able to accommodate a five-fold estimated increase in waste products.

Page | 24

A study conducted by the Nuclear Energy Agency gives an insight on the

increased waste levels that are to be expected with LEU-based targets [26]. For example,

the annual amount of waste corresponding to the 12000 six-day Curies per week

production rate were 215 kg LEU uranium waste vs. 43 kg of HEU waste and 25 g of

Pu-239 for LEU vs. 1.2 g for HEU. These numbers are estimated annually for 20%

enriched targets vs. 93% enriched targets, respectively. It is further noted from the OPAL

reactor in Australia, LEU processed targets will yield increased volumes of intermediate

and low-level liquid waste in comparison to HEU strategies [26]. However, the

possibility remains that by converting liquid waste products to its solid form will

optimize waste storage tanks that are currently operating below capacity and hence,

waste storage will not be significantly impacted [8].

Page | 25

Figure 1-6: Generic Flow Chart of Waste Streams Generated from

Molybdenum-99 Production6

6 Adapted from Reference [27]

Page | 26

1.8 CANDU Reactors and The DRAGON Lattice Code

1.8.1 Using CANDU Reactors for Molybdenum-99 Production

CANDU reactors are known as pressurized heavy water reactors. The acronym

‘CANada Deuterium Uranium’ (CANDU) comes from the reactor’s use of heavy water as

a moderator and NU as fuel. The moderator fills a large tank called the calandria. The

calandria is penetrated by hundreds of horizontal fuel channels, each consisting of an

inner pressure tubes which contains natural uranium oxide fuel in a form of fuel bundles

and an outer calandria tube. The fuel is cooled by heavy water under high pressure. The

pressure channel design makes it possible to refuel the reactor continuously without

shutting it down. This is done by connecting in-tandem both sides of an individual

channel to a fuelling machine and a fuelling machine cooling circuit. Table 1-3 gives a

breakdown of the CANDU reactors and CANDU-derivatives that exist around the world.

Figure 1-7 depicts the important components of a typical CANDU reactor. A

typical CANDU fuel channel is shown in Figure 1-8. The diagram shows the channel

inlet/outlet end fitting and the arrangement of the fuel bundles in the pressure tube.

Page | 27

Table 1-3: Status of CANDU reactors and CANDU-Derivatives in 2014 ([28],

[29])

Country No. of Reactors

Status Capacity (MWe net)

Argentina 3 Operational 1627 Canada 19

6 Operational Permanent Shutdown

13500 2143

China 2 Operational 1300 India 18

4 Operational Under Construction

4091 2520

Republic of Korea

4 Operational 2684

Pakistan 1 Operational 90 Romania 2 Operational 1300

Page | 28

Figure 1-7: CANDU Reactor Face7

7 Reproduced from the CANTEACH website [30]

Page | 29

Figure 1-8: A Typical CANDU Fuel Channel8

8 Reproduced from the CANTEACH website [31]

Page | 30

1.8.2 The DRAGON Code

Analysis of the proposed bundle designs is performed at the lattice level using the

lattice transport code DRAGON version 3.05 [32], [33], [34]. DRAGON has been

developed at École Polytechnique9 de Montreal and is considered one of the

recommended codes by the nuclear industry to perform lattice calculations. The code

solves the integral neutron transport equation using 2-D and 3-D collision probability

techniques.

DRAGON is a lattice code which is used to solve the neutron transport equation

for a given geometry. DRAGON includes all of the functions that characterize a lattice

cell code divided into several calculation modules. The modules are linked together by

GAN generalized driver. The exchange of information is ensured by well-defined data

structures. This is done so that new calculation techniques can be implemented in the

code [35].

DRAGON consists of two main components, one -group collision probability

tracking modules and multigroup flux solvers. The interface-current technique is used

by the JPM tracking option at the level of each homogenous region associated with a

9 The École Polytechnique de Montréal is an engineering school/faculty affiliated with the

University of Montreal in Montreal, Quebec, Canada.

Page | 31

specific geometry. In cases where a multiplicative medium is analyzed, DRAGON

performs power iterations to find the effective multiplication factor.

The effective multiplication factor (keff) is defined as the ratio of the number of

fissions (or fission neutrons) in one generation divided by the number of fissions (or

fission neutrons) in the preceding generation. keff provides an idea of deviation of the

system from equilibrium. The system can be termed critical (the production and

destruction rates are equal), subcritical (the production rate is smaller than destruction

rate) or supercritical (the production rate is greater than destruction rate) based on the

following definition:

𝑘𝑒𝑓𝑓 {< 1 𝑟𝑒𝑎𝑐𝑡𝑜𝑟 𝑠𝑢𝑏𝑐𝑟𝑖𝑡𝑖𝑐𝑎𝑙 = 1 𝑟𝑒𝑎𝑐𝑡𝑜𝑟𝑐𝑟𝑖𝑐𝑡𝑖𝑐𝑎𝑙 > 1 𝑟𝑒𝑎𝑐𝑡𝑜𝑟 𝑠𝑢𝑝𝑒𝑟𝑐𝑟𝑖𝑡𝑖𝑐𝑎𝑙

(1.4)

The multiplication factor of an infinite lattice calculation is represented as kinf. In

the case where the coefficient of reflection at the surface boundary is set to 1

everywhere, as is the case with the models analyzed in this thesis, the outgoing fluxes

will be reflected back into the volume, making the neutron leakage equal to zero, and

thus making keff equal to kinf. It is referred to as the reflective boundary condition.

Another important quantity in reactor neutronics is known as reactivity (ρ). It is

defined as the fractional departure of a system from criticality and is calculated as:

eff

eff

k

k 1 (1.5)

Page | 32

The system reactivity change can be positive or negative depending on the

process involved. For example, addition of any form of poison, like boron in the

moderator or insertion of control rods, brings about a reduction in reactivity, while

removal of poison or addition of fresh fuel or enrichment of NU can bring about an

increase in reactivity.

In the present study, the focus is on change in reactivity from introducing a region

of enriched fuel by comparing it to that of NU. The change in reactivity ∆ρ, introduced

due to enrichment can be defined as:

mkxkk enriched

eff

natrual

eff

100011

(1.6)

There are microscopic cross-section libraries that are structured according to a

standard format and are accessible to DRAGON [34]. The output data stream provided

by DRAGON is the source for obtaining useful information such as macroscopic cross-

section, neutron fluxes, etc.

Page | 33

Chapter 2

Production of Molybdenum-99 in Nuclear Reactors: Present

and Future

Mo-99 is obtained from the fission of U-235, and is principally produced through

the 235U (n,f) 99Mo 99mTc reaction using the five government-owned and funded multi-

purpose reactors listed in Table 1-1. The ageing of these five research reactors has

required ever growing shutdown periods for maintenance and repair. This has made the

supply of Mo-99 more fragile and a topic of concern for the medical community.

Additional challenges are cast on the supply of Mo-99 due to the fact that the

major Mo-99 manufacturers have been using HEU (~93%) for target material, except for

SAFARI-1 in South Africa that uses 45% [8]. Some of the regional producers of Mo-99

are using LEU targets, including for example, the RA-3 reactor in Argentina and the OPAL

reactor in Australia [8].

The challenges arising from the limited number of ageing reactors and the use of

HEU, which ultimately affect the reliable supply of Mo-99 worldwide, will be explored

further in this Chapter.

Page | 34

2.1 Current Molybdenum-99 Supply Challenges

Five nuclear research reactors are producing most of the worldwide supply of

Mo-99 as listed in Table 1-1. In 2007, major disruptions occurred in the global supply of

Mo-99 due to an extended shutdown of the NRU reactor. The disruptions lasted for

several weeks. As a result, many of the diagnostic testing and procedures for life-

threatening conditions were cancelled affecting tens of thousands of patients throughout

Canada and the US [2].

The crisis began when the NRU reactor did not return to operation for several

weeks following a routine maintenance shutdown due to regulatory issues. The NRU

shutdown event was followed by an extension of the planned outage at HFR-Petten due

to leaks, and an unplanned shutdown of an isotope processing facility due to Iodine-131

release from the waste stream [12]. Concerns over the long term supply of medical

isotopes were further exacerbated by the government of Canada’s decision to stop the

development of the MAPLE reactors. These reactors were being designed and built to

meet the world’s demand of medical isotopes, specifically Mo-99, I-125, I-131, and

Xe-133 [36]. There are currently no reliable long-term substitutes to NRU to ensure

continuity in the supply of Mo-99 to the world.

Page | 35

2.2 Conversion of Molybdenum-99 Targets from Highly-Enriched

Uranium to Low-Enriched Uranium

The current producers of Mo-99 use HEU targets of ~93 wt% U-235 with the

exception of SAFARI-1 reactor in South Africa which uses 45 wt% U-235 targets. For the

NRU reactor, HEU from Y-12 National Security Complex in Oak Ridge, Tennessee is

shipped to Chalk River Laboratories in Chalk River, Canada to be manufactured into

Mo-99 targets. Published records by the United States Nuclear Regulatory Commission

(US-NRC) report shipments of 10-25 kg of HEU per year from the US to Canada for Mo-99

production ([20]).

The most pressing change in the medical isotopes industry has been surrounding

the efforts to convert the fuel and the Mo-99 targets to LEU (less than 20% U-235 by

weight). In the interest of nuclear security and non-proliferation, many countries

around the world are making strides in the efforts to migrate from using HEU to LEU.

The movement was supported by the US House of Representatives in November of 2009

in the form of passing a legislation which would eliminate the US export of HEU for

isotope production in Canada within a period of 7 to 10 years [37].

Substituting LEU directly in place of HEU targets would reduce the Mo-99 yield

to approximately 20% of that generated from the HEU due to reduced number of U-235

atoms available for fission. The lower yield can be overcome by irradiating five times

more targets in the reactor, however the available reactor space limitations may

preclude such an option.

Page | 36

Irradiating five times more targets will also result in five times the volume of

separation waste leading to storage and disposal issues. In addition, employing LEU

targets will also increase the generation of fissile plutonium-239 due to neutron capture

by the higher proportion of uranium-238 atoms. However, plutonium can be ultimately

used to make mixed-oxide fuel for nuclear power plants as it is generally insoluble and

can therefore be separated from liquid wastes during processing of the LEU targets [8].

Research is underway to yield more efficient LEU target designs. Researchers are

looking at ways to alter the composition of LEU targets for the purpose of increasing the

density of U-235 in the targets [8]. The Reduced Enrichment for Research and Test

Reactors program was first established in 1978 at the Argonne National Laboratory by

the US Department of Energy. The primary objective of this program was to develop the

technology needed to use LEU instead of HEU in the research and test reactors, without

impacting the experimental performance, economics, or safety of the reactors [38].

There will have to be substantial economic incentives for any large producer of

Mo-99 to change from using HEU. Large investments would be required for

infrastructure, approval of new processes and procedures and the required research.

Any new process would need to be qualified by the various regulators, which is a long

and expensive process, adding to its cost. Implications regarding waste management

with respect to treatment of irradiated targets and disposal should also receive high

priority [39].

Page | 37

2.3 New Producers of Molybdenum-99 Worldwide

The current fleet of Mo-99 producers consisting of the big-five reactors is aging

and is expected to stop the production of Mo-99 in the next decade. The NRU is

scheduled for shutdown and to cease all production in the year 2016. The OSIRIS reactor

is expected to shut down by the end of 2018. These two reactors account for over a

quarter of the available capacity of Mo-99. This will drastically reduce the supply of

Mo-99.

Since the 2007 disruptions in the supply of Mo-99, several countries have

expressed interest in building modern research reactors dedicated to Mo-99 production

in order to curtail reliance on other countries for medical isotopes and to develop

domestic production capabilities. Some countries have proposed new projects to initiate

molybdenum production from an existing facility not originally built for that purpose.

Commissioning a new reactor for isotope production is a lengthy process. According to

a European Commission study on the supply of medical isotopes, it can take ten or more

years for any new project from the planning phase to become operational. If an existing

facility is upgraded to supply Mo-99, these facilities can take five or more years to

become operational depending on available equipment and facilities at the specific

reactor site [7].

Appendix A provides a list of potential producers that have the technological

resources to develop Mo-99 production capabilities. A general underlying objective of

these new initiatives is to develop domestic capabilities of producing Mo-99 and

reducing the dependence on foreign suppliers. In few cases where potential exists to

Page | 38

irradiate a large number of targets, exports of medical isotopes may occur from those

facilities to aid in meeting the global demand for medical isotopes.

2.4 Molybdenum-99 Production in Power Reactors

Production of Mo-99 using commercial power reactors is a novel idea considering

power plants are not normally used to produce radioisotopes. Other radioisotopes have

been produced using power reactors, particularly in CANDU reactors. In Canada, tritium

is recovered from the heavy water coolant [28]. The adjustor rods used in CANDU

reactors are also processed to produce cobalt-60 sources for industrial radiation

processing. However, most commercial power reactors generally restrict their use

solely to the generation of electricity.

The production of medical isotopes in power reactors involves several

interrelated areas such as target fabrication, neutronic analysis, thermalhydraulic

calculations, safety analysis, fuelling schemes, processing and transportation of

irradiated targets [28]. Another important factor is the number and volume of targets

being irradiated at a time, which will determine the yield of radioisotopes. Irradiation

time in reactor is another important factor in obtaining high specific activity.

The CANDU reactors and CANDU-derivatives have a pressure tube design that

permits these reactors to be refuelled online by connecting in-tandem to a fueling

machine. From the standpoint of Mo-99, a case can be made that irradiation of targets

Page | 39

in power reactors is similar to their irradiation in a research reactor in the sense that

these targets can be recovered at any time.

It is also possible to irradiate targets in the irradiation channels that exist in some

pressure tube type reactors ([26], 28]). Instrumentation channels can also be used to

irradiate targets in most reactors. However, this may be a challenging task requiring

reconfiguration of the reactor instrument calibration and possibly the core neutronics.

Irradiation of uranium targets in CANDU has already been proposed in recent

past ([22], [40], [41]). Researchers at McMaster University have analyzed a new 37-

element fuel bundle containing a combination of slightly enriched/depleted uranium

pins that, when irradiated in a CANDU reactor, could yield 100% to 120% of the world

demand of Mo-99 ([22], [40]). The researchers have proposed modifications to the 37-

element fuel bundle that would contain 1.5% to 3% enriched uranium oxide fuel either

in the second and third ring, or in the fourth ring of the fuel bundle, with the rest of fuel

pins containing depleted uranium. Calculations done in the WIMS-AECL-v3.121 lattice

code showed that these targets, when irradiated in the central and periphery channels,

produced significant amounts of Mo-99 without adversely affecting important reactor

safety parameters such as neutronics, heat rating limits, and including any significant

impact on standard fuelling operations. However, the design constraints used in the

McMaster research had wide margins allowing for the Mo-99 producing bundle

neutronics to deviate significantly from the reference value. For example, a reactivity

difference of 10 to 15 mk from the reference CANDU bundle was deemed acceptable for

Page | 40

the Mo-99 producing bundles, potentially leaving a gap in the safety analysis and a need

for confirmation of the design using full core calculations and experimental testing.

Page | 41

Chapter 3

Methodology for the Development of a Molybdenum-99

Producing 37-Element Fuel Bundle Design

3.1 Design Objectives

One of the attractive features of the CANDU reactor is its versatility in fuelling

applications. Opportunity exists for evolutionary improvements in the CANDU fuel

design that can produce large quantities of medical isotopes while maintaining its power

production targets. The ability to refuel online while maintaining a good thermal

neutron economy in the core makes CANDU reactors a strong candidate for producing

significant quantities of Mo-99 to meet the global demand of medical isotopes. On-power

fuelling of the CANDU reactors allows the gradual introduction of new fuel bundle

designs intended for Mo-99 production.

The objective of this thesis is the development of new 37-element fuel bundles

that, when irradiated in a standard CANDU reactor, are able to produce significant

quantities of Mo-99, and which have neutronic and thermalhydraulic properties nearly

identical to those of the standard CANDU fuel bundle. This objective translates into six

design requirements:

1. The bundle shall use LEU fuel of less than 20% U-235 enrichment by

weight and produce Mo-99 in sufficient concentrations to be extracted

using chemical processes currently in use.

Page | 42

2. The hydraulic resistance of the Mo-99-producing bundle shall be identical

to that of a standard CANDU fuel bundle.

3. The bundle shall be dimensionally compatible with the fuel channel

components and with the appropriate fuel handling components.

4. The reactivity of the Mo-99-producing bundle shall be close to that of a

standard CANDU fuel bundle. A reactivity change of ±1 mk of standard

CANDU is deemed acceptable.

5. The power generation rate in the Mo-99-producing bundle shall be similar

to that of the standard CANDU fuel bundle when placed in the same

position in the core.

6. The bundle shall demonstrate that the operating and safety margins

regarding fuel sheath dryout power and fuel centreline temperatures are

maintained with the use of the new bundles in the core.

3.2 Design Plan and Strategy

The plan is to consider various design options that utilize the standard

37-element fuel bundle geometry. The proposed designs for Mo-99 production range

from a fuel bundle comprised of a combination of enriched and depleted UO2 concentric

fuel layers in the outer rings and NU in the centre and inner rings, to a fuel bundle with

mixed fuel pins (enriched + depleted uranium) in all four rings of the 37-element bundle

Page | 43

for maximum yield of Mo-99. Other design options include a fuel-bundle comprised of

uranium metal fuel, which is utilized to achieve higher margins for fuel centreline

melting (due to the metal’s higher thermal conductivity compared to the ceramic) and

ease of manufacturing with respect to the proposed thickness of the enriched fuel region.

Five different design options are assessed in this study. Designs 1, 2 and 3 use

mixed fuel targets (enriched + depleted) in Ring 4 and natural UO2 in the centre and inner

rings. Designs 4 and 5 use mixed fuel targets in all rings. Design 3 and 5 use uranium

metal fuel in the target pins as a test for improved safety margins with respect to fuel

centreline temperatures. A pictorial representation of the various designs options that

are considered in this study is shown in Figure 3-1.

As discussed in Chapter 2, nuclear proliferation concerns have limited the

enrichment in target bundles to 20 wt% of U-235. Due to the fact the ratio of Mo-99 yield

to the total fuel volume processed is low for NU, from the perspective of maximizing the

yields of Mo-99, use of higher enrichment is advantageous. Hence, use of 19.5 wt% of

U-235 enrichment, which is just below the HEU category, is recommended for this study.

The depleted uranium region is assumed to have 0.20% of U-235 by weight.

Depleted uranium is a by-product of the uranium enrichment process. Different fuel

enrichment facilities may have different end-cycle enrichment of uranium tails. Because

the enrichment of most uranium tails is typically less than 0.3% [42], 0.2 % is an easily

achievable value to be used as depleted uranium fuel in the proposed designs.

Page | 44

Figure 3-1: A Pictorial Representation of the Proposed Designs Assessed for

Molybdenum-99 Production10

10 The base image of the 37-elment bundle is adapted from Reference [43]

19.5% Enriched

UO2

0.2% Depleted

UO2

Natural UO2

19.5% Enriched

UO2

0.2% Depleted

UO2

Natural UO2

19.5% Enriched

U-Metal

0.2% Depleted

U-Metal

Natural UO2

19.5% Enriched

UO2

0.2% Depleted

UO2

19.5% Enriched

U-Metal

0.2% Depleted

U-Metal

Page | 45

3.2.1 Satisfying Design Requirements

In this section, the methodology that is adopted to satisfy the design

requirements listed in Section 3.1 is presented in detail.

Requirements 2 and 3:

The new fuel bundle design is based on the standard 37-element CANDU fuel

bundle. The bundle geometry is kept dimensionally identical to the standard CANDU

bundle. This allows the new design to have the same hydraulic resistance as the

reference bundle and, hence, be compatible with fuel channel components and the

appropriate fuel handling components in order to allow online refuelling.

Requirements 2 and 3 are thus satisfied.

Requirement 4:

Fuel pins in some (or all, depending on the specific design) rings are divided into

two concentric regions, one consisting of enriched uranium fuel and another consisting

of depleted uranium fuel. The relative dimension and position of the enriched-fuel and

depleted-fuel regions are determined from the condition that the bundle has a fresh fuel

infinite-lattice k-infinity (kinf) value that is similar to the standard CANDU fuel lattice

along with the condition on the fuel heating and centreline temperature limits. The cell

fresh fuel kinf is determined by the DRAGON code for the specified infinite lattice model

geometry. The kinf value needs to be as close as possible to the reference CANDU bundle

to allow for existing operational practices and bundle safety characteristics to be

maintained.

Page | 46

In order to meet the bundle reactivity criterion (Requirement 4), Task 1 has been

created and addressed further in Section 4.1.

Requirement 5:

Requirement 5 is implicit in the sense that a bundle with the same kinf as the

standard CANDU bundle and with identical overall bundle geometry will have average

two-group cross-sections similar to the standard CANDU bundle resulting in similar

power generation rates. Hence, Requirement 5 is also met by satisfying Requirements 2

and 4.

Task 1

Calculate the thickness (or radius) of enriched fuel region for the various

design options considered which yields a fresh fuel infinite-lattice kinf

value that is similar to that of the reference CANDU bundle. For this

study, a difference in reactivity (Δρ) value of ±1 mk compared to the

reference CANDU bundle is considered acceptable.

Page | 47

Requirement 6:

Fuel sheath dryout and fuel centreline temperatures are two phenomena that

serve as indicators of the onset of a heat transfer crisis and deteriorating bundle safety

characteristics. Heat generated in the fuel volume must be transported out through the

fuel surface.

The linear power rating represents the power produced per unit length (W/m)

of a fuel element. The linear heat rating of the element primarily dictates the sheath

temperature rise in case of a dryout. The heat flux is defined as the heat transferred per

unit surface area (W/m2). Generating excessive power from a bundle causes sheath

temperature to rise and bundle critical heat flux (CHF) and fuel centreline melting

margins to diminish. The CHF characteristics of a bundle are functions of bundle radial

flux distribution from ring-to-ring, bearing pad height, channel creep profile, fuel string

axial flux distribution. Other design features (such as end cap, end plate, etc.) also affect

CHF characteristics of a bundle.

Designs 1, 2 and 4 have the same fuel pellet and sheath diameters as the reference

CANDU bundle. This ensures the ability of the bundle to maintain the same

hydrodynamic properties as the reference CANDU bundle. For a bundle that is

dimensionally identical to the reference bundle with similar power generation rates, its

available margins to CHF can be determined by making a comparison of the linear power

ratings of the highest-power pins in both bundles.

Page | 48

Designs 3 and 5, which use uranium metal as fuel, have the same pin diameter,

but a smaller fuel pellet diameter. The increase in parasitic absorption due to the thicker

cladding is balanced by increasing the thickness of the enriched layer as part of the

activities performed under Task 1. Similar assertions regarding the CHF margins can be

made for Designs 3 and 5.

The second important safety parameter is fuel centreline temperature, which

must not exceed the values specified for the standard CANDU bundle for the same bundle

heat generation rates. The new fuel bundles will have non-homogenous fuel

composition containing concentric fuel regions inside Mo-99 target pins. Hence, meeting

the linear heat rating limits alone will not ensure adequate margins to the fuel centreline

melting.

The outermost ring (Ring 4) of the 37-element fuel bundle generally has the

highest linear heat rating value due to its proximity to the moderator and therefore to

the highest flux of thermalized neutrons. This is especially true if the outermost ring

contains a region of enriched fuel as is the case in the proposed designs (see Figure 3-1).

The enriched fuel region, due to its increased rate of fission and consequently higher

power density, creates the possibility of having a higher than reference centreline

temperature in the fuel.

To satisfy Requirement 6 regarding bundle safety characteristics, Task 2 and

Task 3 are created and addressed further in Sections 4.2 and 4.3.

Page | 49

Task 2

The modified bundle designs need to comply with the existing linear heat

rating, bundle power and channel power limits prescribed for standard

CANDU fuel. Specifically, the new designs must maintain a maximum

linear heat rating of less than 60 kW/m [44]. Ideally, the bundle must

have a ring power distribution that is similar to the existing bundles.

Task 3

Produce a radial temperature profile of the fuel pin with the highest

linear heat rating and compare it with the radial temperature profile of

the reference fuel pin of the same ring to ensure that adequate margins

to fuel centreline melting are maintained. The melting points of UO2

(2860 °C) and U-metal (1132 °C) are used as constraints here.

Page | 50

Requirement 1:

The enrichment of the enriched-fuel region is fixed at 19.5 wt% of U-235 which

is sufficiently high to allow Mo-99 extraction by existing processes while being slightly

below the 20 wt% limit for low-enriched fuel. According to an IAEA report [45], fuel

reprocessing technologies exist for both LEU metal and UO2 targets for Mo-99

production. Beside some modification needed to accommodate the larger masses of LEU

target material, the rest of the chemical processing involved for the production of Mo-99

using LEU targets is very similar to the present process for producing Mo-99 using HEU.

The expert committee report by the US National Academy of Sciences also

concluded that the there are no technical barriers to producing large quantities of Mo-99

from LEU targets [8].

For the modified fuel bundles proposed in this thesis, the Mo-99 producing fuel

pins are expected to be chemically processed using the modified acid dissolution

technique. After irradiation, the enriched fuel pellets are removed from the cladding and

processed for Mo-99 extraction. The mass and activity of Mo-99 that is expected to be

produced in these pins can be calculated from the output of the DRAGON code. The code

output is obtained for each specified lattice sub-region of the modified fuel bundle

geometrical model. The mass and activity of Mo-99 produced is also explored as part of

this thesis in Section 4.4. The expected yield of Mo-99 in six-day Curies is calculated for

each new bundle design.

Page | 51

3.3 Reactor Operating Conditions and Assumptions

The modified fuel bundles with different geometrical arrangements of 19.5 wt%

enriched, 0.2 wt% depleted and 0.71 wt% uranium fuels in different fuel rings are

subject to the same conditions and power limits as the existing fuel bundles in an

operating CANDU core. The new fuel bundle is designed to withstand irradiation in the

highest power channel of the 380-channel CANDU-6 reactor with a total thermal power

of approximately 2x109 Wth. A bundle power of 900 kW is used as the upper limit for a

typical CANDU fuel assembly in order to meet the existing channel power limits. The

maximum licensed bundle power limit for a CANDU-6 fuel-channel using 37-element fuel

bundle is 935 kW. This corresponds to a linear power rating in the hottest (outermost)

fuel element of approximately 60 kW/m.

Therefore, for all new designs considered, the lattice cell bundle power at

100% FP is assumed to correspond to the maximum instantaneous power of 900 kW.

The reference CANDU-6 fuel bundle contains 19.85 kg of uranium fuel. Hence, the

corresponding bundle specific power is 45.34 W/g (900 kW/19.85 kg per bundle). In

order to assess the safety margins of the proposed designs, the pin powers and linear

ratings are scaled according to the maximum operating CANDU-6 bundle specific power.

The bundle designs are analyzed using the reactor physics code DRAGON

version 3.05 ([32], [33], [34]). Transport calculations are performed using the collision

probability method as implemented in the DRAGON code. The model used for this study

is a single-cell infinite lattice with reflective boundaries on all sides. The effects of

reactivity devices are ignored in the single-cell DRAGON model.

Page | 52

The heat generated by the bundle is conducted through the fuel rod and

convected by the surrounding coolant in the fuel channel. In the current study, the fuel

temperature is calculated for the fuel pin with the highest linear heat rating at various

points along the radius of the pin, including the centre. The fuel temperature profile of

a 4th ring pin is calculated using a numerical integration method (see Section 3.4.2) for

each design and compared to the fuel melting limits.

The numerical integration method assumes a radial conduction model for a slice

of fuel element based on steady-state conditions where the fuel element geometry,

thermal properties, and physical characteristics are known. Axial conduction is assumed

negligible and thermal conductivity of the cladding material is assumed constant.

The melting points of UO2 (2860 °C) and U-metal (1132 °C) are used as the

maximum allowable fuel centreline temperatures. Steady-state cooling flow conditions

are assumed in the fuel channel resulting in a constant bulk coolant temperature.

Temperature of the coolant is assumed to be 276 °C, which is 10 °C higher than the

average CANDU-6 inlet temperature [46], added for conservatism in the analysis. The

difference in the proposed and reference CANDU temperature profiles and the available

margins to fuel melting limits provide high confidence in the adopted designs.

3.3.1 Reactor Fuelling Considerations

The use of modified fuel bundles for the production of Mo-99 will have an impact

on the regular refuelling cycles at CANDU stations. As previously explained, irradiation

times of about 15 to 20 days are sufficient to achieve maximum possible activity of

Page | 53

Mo-99. A burnup of 906.62 MWD/T, which corresponds to 20 days of irradiation at the

specific power of 45.33 kW/kg is used in this study.

The burnup time required to reach the maximum activity levels of Mo-99 is much

shorter than the normal time a regular fuel bundle spends inside the reactor (8 to 12

months depending on the location of the bundle in the core). As a consequence of the

shorter irradiation period, a fuelling machine visit is needed after the peak production

rate for Mo-99 is reached. To achieve steady production levels of Mo-99, two or more

modified fuel bundles can be loaded together at a time. By staggering the production in

multiple channels, a steady amount of Mo-99 producing bundles can be removed each

day to provide a constant yield.

Researchers from McMaster University have studied the impact of shorter

irradiation cycles for Mo-99 producing bundles on the standard fuelling operations [22].

This involved changing the total number of fuelling machine visits that may be required

in order to produce significant yields of Mo-99. The study simulated two different

fuelling scenarios consisting of 3 to 6 fuel channels containing 2 to 6 Mo-99 target

bundles in a staggered manner. The study concluded that with an additional 6 to 8

fuelling operations per week, a production level of 4 bundles per week is attainable [22].

Due to the unique flexibility of fuelling arrangement in CANDU reactors, a yield of

higher than 4 bundles per week is also possible. For the new designs proposed in this

study, the consequent changes in the existing fuelling schemes are not studied as no

difficulties are foreseen. The number of bundles to be extracted per week is assumed to

Page | 54

be a function of Mo-99 yield per bundle and the underlying commercial and medical

isotope market considerations.

Once an approximate activity of Mo-99 per bundle is known, a detailed

calculation based on the instantaneous channel parameters can be performed at various

stages in the fuelling cycle. The bundle fuelling scheme can then be optimized to

determine the most favourable irradiation times for each stage and bundle position for

the Mo-99 production bundles. Using multiple channels in a staggered manner would

ensure a yield of multiple bundles for a constant output of Mo-99 required for the target

medical isotopes market.

3.4 Design Options and Methodology

To develop a new 37-element fuel bundle aimed at producing large quantities of

Mo-99, five different design options are assessed. They include:

Page | 55

Design 1

19.5 wt% enriched UO2 fuel in the centre of the 4th ring fuel elements

encapsulated by 0.2 wt% depleted UO2 fuel. Fuel elements in the centre

and inner rings contain natural UO2.

Design 2

0.2 wt% depleted UO2 fuel in the centre of the 4th ring fuel elements

encapsulated by 19.5 wt% enriched UO2 fuel. Fuel elements in the centre

and inner rings contain natural UO2.

Design 3

0.2 wt% depleted U-metal fuel in the centre of the 4th ring fuel elements

encapsulated by 19.5 wt% enriched U-metal fuel. Fuel elements in the

centre and inner rings contain natural UO2.

Design 4

0.2 wt% depleted UO2 fuel in the centre of the fuel elements encapsulated

by 19.5 wt% enriched UO2 fuel in all four rings.

Design 5

0.2 wt% depleted U-metal fuel in the centre of the fuel elements

encapsulated by 19.5 wt% enriched U-metal fuel in all four rings.

Page | 56

A unit lattice cell is analyzed using the DRAGON neutron-transport code. The cell

is initially divided into 37 homogenous regions to represent different parts of the fuel

bundle including the four fuel rings, cladding, annulus gas, pressure tube and coolant.

Depending on the design, the regions are further split into sub-regions for accuracy and

higher resolution. For example, in Design 2 which contains an enriched uranium fuel

region encapsulated by depleted uranium fuel in the 4th ring and NU in other rings (see

Figure 3-1), the depleted and enriched fuel regions in the fuel pellet are further split into

10 and 3 sub-regions, respectively. In the remaining rings, a total of 4 sub-regions are

used to simulate the NU fuel pellet. Figure 3-2 illustrates the 46 (37 +9 additional

splitting) lattice sub-regions of the DRAGON geometrical model that is used to represent

the 37-element fuel bundle.

The multigroup transport equation is solved using the DRAGON code and the

69-group IAEA WIMS-D Library Update (WLUP) [34] for a single-cell lattice model to

find the neutron flux and cross-sections in each of these sub-regions integrated over all

energy groups. The fresh fuel kinf values are recorded for each case for assessment.

Data on lattice specifications for a 37-element CANDU fuel bundle is provided in

Appendix B.

Page | 57

Figure 3-2 37-Element CANDU Fuel Bundle and the Infinite Lattice Cell

Representation of Design 211

3.4.1 Method for Calculating the Linear Heat Rating

For the purpose of establishing the bounding linear heat rating for fuel pins of the

proposed bundles for Mo-99 production, this study assumes that a fresh target bundle

will be loaded in the middle of the highest power fuel channel facing a high thermal-

neutron flux and irradiated there for the duration of the irradiation period. In other

words, the position of the fuel bundle is assumed to be in position 6 or 7 out of a total of

12 fuel bundles in each CANDU fuel channel. This is a rather conservative case

considering the bundle would initially be loaded anywhere between position 1 to 12

11 The 37-element fuel bundle image is reproduced from the CANTEACH website [31]

Page | 58

depending on the fuelling scheme adopted, and hence could initially have a lower power

rating.

The maximum instantaneous bundle power located in the centre of the CANDU

core is 900 kWth. The pin power ratings are scaled according to the reference CANDU-6

bundle specific power as listed in Section 3.3. The procedure used to evaluate the linear

heat rating is provided within this section.

First, the density of uranium dv, U is computed, followed by the calculation of the

total linear density dL, U of the initial heavy elements in the fuel.

2

2v,

,v

UO

UUO

UA

Add (3.1)

regions

iiDUUL Vdd

1

2,v, (3.2)

where dv, UO2 is the density of natural uranium oxide fuel, AU is atomic mass of

natural uranium, AUO2 is atomic mass of uranium oxide, and V2-Di is the 2-D volume (cm2)

of each fuel region in the lattice cell.

The normalized linear heat rating QL’ for a fuel bundle is then given by

bundleULL PdQ ,

' (3.3)

where Pbundle is the normalized bundle specific power which is equal to 45.34 W/g

(based on 900 kW maximum bundle power), used for scaling with the reference

CANDU-6 bundle.

Page | 59

The unnormalized linear heat rating qi,u’ for each fuel region ‘i’ is given by:

'

, , 2ˆ

i u i f i i Dq e V (3.4)

where ei is the energy produced per fission, Σf,i is the macroscopic fission cross-

section at burnup, and ̂ is the condensed neutron energy group volume-integrated flux

at each region at burnup.

The total unnormalized linear heat rating QL,u’ for the entire fuel bundle is

therefore:

regions

i

uiuL qQ1

'

,

'

, (3.5)

The normalized linear heat power and the unnormalized linear heat rating for a

region within a fuel element are related by the following formula:

ui

uL

Li q

Q

QQ ,

'

'

,

'

'

(3.6)

In DRAGON, the multiple sub-regions (or splits) within one fuel type (enriched,

depleted, or natural) can be merged to provide cross-sections and fluxes for each fuel

type using volume averaging. The bundle length is also constant.

The linear power rating is important for fuel pins, but not for sub-regions inside

each fuel pin. Therefore, once the linear power in each fuel type within one fuel element

is known, the overall linear power for a fuel element composed of different fuel regions

(enriched + depleted) is the summation of two linear powers.

Page | 60

Finally, since there are rpinsN , pins associated with a ring, the average pin heat

rating, '

rQ , in ring ‘r’ can be computed by the following formula:

rpins

regionfueldepletedregionfuelenriched

rN

QQQ

,

''' (3.7)

3.4.2 Method for Calculating the Radial Temperature Profile

The heat generated in a fuel pin directly affects the fuel centreline temperature

and the temperature profile going from the fuel centre to the fuel surface. The fuel

centreline temperature and the temperature gradient are functions of thermal

conduction within the fuel pellet, heat transfer through the diametral fuel-sheath gap,

and conduction through the sheath, coolant temperature and sheath-to-coolant heat

transfer coefficient.

The primary heat transfer mechanisms within the diametral gap are conduction

through the fill gas and solid-to-solid contact conduction (fuel and sheath are in contact).

For the purpose of this work, the fuel pellet is assumed to remain in contact with the

cladding making the thickness of the gap between the fuel pellet and the cladding equal

to zero. This is a reasonable assumption considering that the Mo-99 producing fuel

bundles will use collapsible sheaths and are expected to be irradiated for a maximum of

20 days, which is not long enough for a gap to develop.

As part of the new fuel bundle development for Mo-99 production, a radial fuel

temperature profile of the 4th ring element of the proposed fuel bundles is produced and

compared against the fuel melting limits. The enriched fuel region in the fuel pellets adds

Page | 61

to the possibility of having higher centreline temperatures in the pin due to the increased

rate of fission and consequently higher power density in the region. Due to its proximity

to a large number of thermalized neutrons, the elements of the 4th ring present the most

conservative estimate for centreline calculations.

For the current study, the fuel centreline temperature is calculated using a

numerical integration method described below.

Numerical Integration Method

For any new fuel design, the interface between the fuel and the coolant is centrally

important to the design of the reactor because it limits the power output. The heat

transfer coefficient (hc) of the coolant is dependent on the fluid properties and flow

conditions. For liquid-to-wall heat transfer, the coefficient can be estimated by the use

of certain dimensionless fluid mechanic numbers:

Reynolds number

Prandtl number

Nusselt number

The Reynolds number is a measure of the relationship between the inertial and

viscous effects of the moving fluid and can be defined as [47]:

hVDRe

(3.8)

Page | 62

where ρ is density of the fluid, V is average fluid velocity, Dh is hydraulic radius of

the flow path, and μ is dynamic viscosity of the fluid.

wet

flow

hP

AD

4

(3.9)

where Aflow is the coolant flow area through the fuel bundle, and Pwet is the wetted

parameter.

The Prandtl number is a measure of the fluid thermal properties and can be

defined as:

k

c pPr

(3.10)

where cp is specific heat of the fluid, μ is dynamic viscosity of the fluid, and k is

thermal conductivity of the fluid. Some software such as NIST REFPROP can also provide

Prandtl number and other useful thermophysical properties of fluids at the specified

fluid state.

For the specific case of convective heat transfer inside a tube, the Nusselt number

is an empirical formula that is defined as:

4.08.0 PrRe023.0Nu (3.11)

where Re is Reynolds number, and Pr is Prandtl number. The relationship

between the Nusselt number and the heat transfer coefficient h can be defined as [48]:

Page | 63

k

hDNu h

(3.12)

For a solid, the general thermal energy balance equation of an arbitrary volume

can be expressed as:

dsntrQdVtrQdVt

e

ˆ),(''),('''

(3.13)

[stored] [generated] [transferred]

where ρ is the material density, e is the internal energy, V is the volume, S is the

surface area, Q''' is the volumetric heat generation, Q'' is the heat flux, and n̂ is the unit

vector on the surface. The internal energy can be replaced with temperature T times the

heat capacity c. Using Gauss law, the heat balance equation can be transformed into:

SV

dsnQQ ˆ'''' (3.14)

),(''),(''')(

trQtrQt

cT

(3.15)

Using Fourier law: ),(),('' trTktrQ , the equation governing the fuel

temperature distribution can be derived.

),(),(''')(

trTktrQt

cT

(3.16)

Page | 64

In order to calculate the fuel centreline temperature, steady-state one-

dimensional heat transfer with volumetric heat source and variation of thermal

conductivity with temperature is assumed.

),(),(''' trTktrQ f (3.17)

1( ( ) ) '''( )f

d dTk T r Q r

r dr dr (3.18)

( ( ) ) '''( )f

dTd k T r Q r rdr

dr (3.19)

Integrating both sides from the centreline to an arbitrary radius r:

0

( ) '''( )

r

f

dTk T r Q d

dr

(3.20)

Since

2

)(')('''

0

rQdQ

r

(3.21)

Therefore,

'( )( )

2f

dT Q rk T r

dr

(3.22)

'( )( )

2f

Q rk T dT dr

r

(3.23)

Page | 65

Integrating both sides again from an arbitrary radius r to the radius of the fuel

pin rf:

1 '( )( )

2

fsrT

f

T r

Qk T dT d

(3.24)

or, equivalently:

1 '( )( )

2

f

s

rT

f

T r

Qk T dT d

(3.25)

where,

35.16exp

6400

6142.3692.175408.7

100)(

2/52Tk f (3.26)

where = T/1000, T is in Kelvin, and kf is thermal conductivity for 95% dense

UO2 in W/m∙K. The above expression for thermal conductivity is taken from the IAEA

materials database [49]. The expression is plotted in Figure 3-3.

Figure 3-3: A Plot of UO2 Thermal Conductivity Showing Variation Due to

Temperature

2.0

2.5

3.0

3.5

4.0

4.5

5.0

5.5

400 600 800 1000 1200 1400 1600 1800

Th

erm

al

Co

nd

uct

ivit

y (

W/

m∙K

)

Temperature (K)

Page | 66

Since the temperature dependence of thermal conductivity kf(T) is known, the

left-hand side of Eq. (3.25) is a function, which can be calculated using the trapezoidal

rule to approximate the integral and equate with the right-hand side. Therefore, the

equation can be re-written as:

( , ) ( )

s

T

s f

T

F T T k T dT (3.27)

Or, equivalently:

1 '( )( , )

2

fr

s

r

QF T T d

(3.28)

The solution of Eq. (3.28) gives the temperature T as a function of radius r

corresponding to different points on the pin. If the radius r is taken to be zero

(integration from the centreline to the outer radius of the pin), then:

0

1 '( )( )

2

fcL

s

rT

f

T

Qk T dT d

(3.29)

Or, equivalently:

0

1 '( )( , )

2

fr

cL s

QF T T d

(3.30)

It is to be noted that the integrand in the right-hand side of Eq. (3.30) vanishes as

the radius approaches zero. In mathematical form:

Page | 67

0 L'Hopital's Rule 0 0

'( )

'( )lim lim lim 2 '''( ) 0

dQ

Q dQ

d

d

(3.31)

For a fuel pin that is divided into n annular regions of radii ri, the temperature at

boundaries between annuli will be denoted by T(ri)=Ti, where rn=rf and r0=0, and for

temperatures, T0=TcL and Tn=Ts. Each annulus will have the same index ‘i’ as its outer

radius.

In what follows, discrete forms of Eq. (3.21), Eq. (3.27) and Eq. (3.28) are

presented. The discrete forms of each equation are obtained by approximating the

integral in the respective equation. Because DRAGON, provides the average heat

generation rate for each annulus, Eq. (3.21) is approximated as:

2 2

1

1

'( )'''

2 2

ij j i

i

j

r r Q rQ

(3.32)

It can be observed that the above approximation provides the values of the linear

heat generation rate at the boundaries between annuli, which is at radii ri. To discretize

Eq. (3.27), kf(T) is assumed to be tabulated at discrete values Tk. using Eq. (3.26), and

with T0=Ts. Using the trapezoidal rule to approximate the integral, Eq. (3.27) becomes:

1

1

1

( ) ( )( , )

2

kf j f j

k s j j

j

k T k TF T T T T

(3.33)

For temperatures between the tabulation points Tk-1 and Tk, the function F can be

interpolated linearly as:

Page | 68

11 1

1

( , ) ( , )( , ) ( , ) k s k s

s k s k

k k

F T T F T TF T T F T T T T

T T

(3.34)

Because the values of the linear heat generation rate Q’ is calculated (according

to Eq. (3.32)) at radii ri, the integral in Eq. (3.28) can be approximated using the

trapezoidal rule:

11

1 1

'( ) '( )1( , )

2 2

nj jj j

i s

j i j j

r rQ r Q rF T T

r r

(3.35)

From a practical perspective, Eq. (3.35) is solved by finding the interval (Tk-1, Tk)

such that:

11

1

1 1

'( ) '( )1( , ) ( , )

2 2

nj jj j

k s k s

j i j j

r rQ r Q rF T T F T T

r r

(3.36)

And then solving:

111

1 1

11 1

'( ) '( )( , ) ( , ) 1( , )

2 2

nj jj jk s k s

k s i k

j ik k j j

r rQ r Q rF T T F T TF T T T T

T T r r

(3.37)

The fuel sheath (cladding) is exposed to constant heat flux. For the purpose of

this study, it is assumed that no gap exists between the fuel and cladding. Also since

Q’’’clad =0, heat conduction in the clad then becomes:

0)(1

dr

dTrk

dr

d

rc

(3.38)

Page | 69

)ln(2

'

)( 1

F

cF

n

i

i

scccladcr

trQ

TTkTk

(3.39)

Heat transfer from the clad to coolant can be represented by Newton’s law of

cooling (by convection).

)(''1

Bss

n

i

i TThQ

(3.40)

)(2

'

)(2

'''''

)( 1

2

11

cGs

n

i

i

cGs

i

n

i

i

s

n

i

i

BsCooltrh

Q

trh

rQ

h

Q

TTT

(3.41)

The phenomenon is dominated by the presence of a thin surface layer. The

temperature drop is assumed to occur over this thin layer and the bulk of the coolant is

assumed to have a constant temperature of 549 K.

3.4.3 Method for Calculating the Mass and Activity of Molybdenum-99

Produced

The IAEA-WLUP library [34] does not include fission yield data for Mo-99 and, as

a consequence, DRAGON cannot explicitly calculate the atomic density (or mass) of

Mo-99 produced from fission. Hence, the mass and activity of Mo-99 must be calculated

based on the total fission rates calculated by DRAGON. The following methodology is

used for calculating the mass and activity of Mo-99 produced.

Rate of change = Rate produced – Decay rate

Page | 70

NVcfdt

dNUenrichedfMo 235,99

ˆ (3.42)

where N is the number of Mo-99 atoms produced as a fission product, 99Mof is

the Mo-99 fission yield, f is the macroscopic fission cross-section at burnup (cm-1), c

is the DRAGON scaling multiplier, ̂ is the condensed neutron volume integrated flux in

the enriched region at burnup (n·cm/s), and 235, UenrichedV is the volume of the enriched

U-235 region in the bundle (cm3). Furthermore, the activation of Mo-99 by capturing a

neutron can be neglected here due to a small cross-section value of this reaction (σabs =

2.65 barns).

Let 235,99ˆ

UenrichedfMoprod VcfR (3.43)

The equation then becomes:

NRdt

dNprod (3.44)

NdtdtRdN prod (3.45)

dtRNdtdN prod (3.46)

Multiply both sides by eλt

dtReNdtedNe prod

ttt (3.47)

dteRNed t

prod

t )( (3.48)

Page | 71

Integrate both sides with-respect-to time

dteRNed

t

t

prod

t

t

00

)( (3.49)

The neutrons in the fuel region are not mono-energetic. There exists a spectrum

of neutron energy distribution. However, an energy condensed volume integrated

neutron flux is assumed here. As well, the corresponding average macroscopic cross-

sections are used. Hence the rate of production in this case is a constant, and Eq. (3.49)

then becomes:

dteRNed

t

t

prod

t

t

00

)( (3.50)

prodtprodtR

eR

NNe 0 (3.51)

Multiply both sides by e-λt

ttprodprodeNe

RRtN

0)( (3.52)

N0 is the initial number of Mo-99 items at t=0, which is equal to zero.

tprode

RtN

1)( (3.53)

Or, equivalently:

t

prod eRttN 1)()( (3.54)

Page | 72

t

UenrichedfMo eVcft

1ˆ)( 235,99 (Bq) (3.55)

where, )(t is activity at irradiation time t. The mass of Mo-99 produced then

becomes:

23

99

10023.6.1)(

x

Ae

Rtm Motprod

(3.56)

23

99235,99

10023.6.1

ˆ)(

x

Ae

Vcftm MotUenrichedfMo

(grams) (3.57)

where )(tm is the mass of Mo-99 produced due at irradiation time t, and 99MoA

is the atomic weight of Mo-99.

Eq. (3.56) and Eq. (3.57) above calculate the end-of-bombardment (EOB) activity

and mass of Mo-99 in fission products. The processing efficiency for Mo-99 is about 90%

and takes approximately one day, resulting in a post processing total that is equivalent

to 70.65% of the EOB value. The six-day Curies value is then computed as:

decayp

t

UenrichedfMo eVcft

1ˆ)( 235,99 (Bq) (3.58)

decayp

MotUenrichedfMo

x

Ae

Vcftm

23

99235,99

10023.6.1

ˆ)( (grams) (3.59)

Where p is the post processing efficient of 70.65%, and

decay is computed using:

22.0)6( )6(

0

days

decay eN

daysN (3.60)

Page | 73

Chapter 4

Assessment Results and Discussion

This chapter provides the results obtained from the DRAGON lattice code along

with the calculations of the linear heat ratings and fuel centreline temperatures for the

proposed designs.

As mentioned before, five different design options are assessed in this study.

Information on these proposed variations of the standard 37-element CANDU bundle

designs for Mo-99 production are provided in Sections 3.2 and 3.4. For each design, the

fuel pellets in the target rings are made up of two-regions, a region of enriched fuel

surrounded by a region of depleted uranium (or vice versa). Enrichment of 19.5 wt%

U-235, which is less than the threshold of HEU, was used in the modified designs. The

depleted uranium was assumed to have 0.20 wt% U-235 enrichment.

A single cell infinite lattice assembly was modelled using reactor physics code

DRAGON version 3.05. The DRAGON model considers a fresh fuel bundle at the start and

irradiates it for 20 days corresponding to a burnup of 906.62 MWD/T. This burn is

sufficient to reach the maximum activity of Mo-99 in the target fuel.

The Mo-99 production bundles are designed to meet the same safety and design

constraints as the standard 37-element CANDU fuel with respect to reactor neutronics

and heat generation in the fuel elements. The standard CANDU bundle with NU fuel is

used as a benchmark for new designs. The reference bundle assessment results are

provided in Tables 4-1 through 4-3 and Figures 4-1 and 4-2.

Page | 74

The DRAGON calculated fresh fuel kinf for the reference CANDU unitcell with NU

fuel bundle and heavy water coolant is found to be 1.1228. The kinf is greater than 1 since

the effects of adjuster rods and reactivity devices are not modelled in the DRAGON

unitcell lattice model. The reactivity devices absorb neutrons and result in the kinf being

closer to 1.

Table 4-1: Reference CANDU 37-Element Bundle Data Structures for Radial

Pin Power Profile Calculations

Region

Macroscopic Fission Cross-Section, Σf (cm-1)

Integrated Flux, ̂

(n·cm/s)

2-D Volume, V2-D

(cm2)

Unnormalized Linear Power Rating, qi.u (kW/cm)

Linear Density, dL, U (g/cm)

Normalized Linear Power

Rating , '

rQ

(kW/m) Ring 1 0.02236 0.3682 1.159 2.696E-16 10.83 37.68 Ring 2 0.02330 2.224 6.957 1.697E-15 65.00 39.53 Ring 3 0.02595 4.534 13.91 3.852E-15 130.0 44.87 Ring 4 0.03091 7.018 20.87 7.181E-15 195.0 55.76

Figure 4-1: Radial Pin Power Profile for Reference CANDU 37-Element Bundle

30

35

40

45

50

55

60

1 2 3 4LIN

EA

R H

EA

T R

AT

ING

(K

W/M

)

FUEL BUNDLE RINGS

Reference CANDU Fuel Bundle

Page | 75

Table 4-2: Reference CANDU 37-Element Bundle Normalized Heat Density and

Linear Power Rating in Ring 4 Sub-regions

Sub-regions in Ring 4

Enrichment (wt%)

Thickness (cm)

Pin Radii (cm)

Normalized Heat Density, Q’’’ (kW/cm3)

Cumulative Normalized Linear Power Rating, Q’ (kW/cm)

Ring4-Disk1 0.7110 0.03757 0.03757 468.5 2.077 Ring4-Disk2 0.7110 0.03757 0.07513 469.6 8.324 Ring4-Disk3 0.7110 0.03757 0.1127 469.7 18.73 Ring4-Disk4 0.7110 0.04948 0.1622 469.9 38.81 Ring4-Disk5 0.7110 0.04948 0.2117 470.9 66.18 Ring4-Disk6 0.7110 0.04938 0.2611 472.1 100.8 Ring4-Disk7 0.7110 0.04938 0.3106 473.5 142.9 Ring4-Disk8 0.7110 0.04948 0.3601 475.4 192.5 Ring4-Disk9 0.7110 0.04938 0.4096 477.6 249.6 Ring4-Disk10 0.7110 0.04938 0.4591 480.4 314.5 Ring4-Disk11 0.7110 0.04948 0.5085 483.7 387.3 Ring4-Disk12 0.7110 0.04938 0.5580 487.8 468.2 Ring4-Disk13 0.7110 0.04948 0.6075 493.5 557.6

Page | 76

Table 4-3: Reference CANDU 37-Element Bundle Numerical Integration

Results Using IAEA and Constant Thermal Conductivity for UO2

Node Temperature using IAEA Thermal Conductivity (°C)

Temperature using Constant Thermal Conductivity (°C)

Fuel Centreline 1919 2197 Node1 (Ring4) 1911 2190 Node2 (Ring4) 1887 2170 Node3 (Ring4) 1847 2135 Node4 (Ring4) 1770 2069 Node5 (Ring4) 1665 1978 Node6 (Ring4) 1533 1863 Node7 (Ring4) 1380 1724 Node8 (Ring4) 1211 1561 Node9 (Ring4) 1036 1373 Node10 (Ring4) 860.9 1161 Node11 (Ring4) 689.9 924.0 Node12 (Ring4) 526.6 661.4 Node13 (Ring4) 373.2 373.2 Node14 (Sheath) 309.7 309.7 Node15 (Coolant) 275.8 275.8

Page | 77

Figure 4-2: Reference CANDU 37-Element Bundle Radial Temperature Profile

of Ring 4 Pin Using IAEA and Constant Thermal Conductivity for UO2

0

500

1000

1500

2000

2500

3000

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8

Te

mp

era

ture

(°C

)

Radial Distance (cm)

IAEA Thermal Conductivity (Eq.3.26)

Constant Thermal Conductivity(0.024 W/cm-K)

UO2 Melting Point

Page | 78

4.1 Task 1 - Comparison of Infinite Multiplicative Constants

Using an infinite lattice cell model, the fresh fuel kinf value of the new designs is

used to evaluate the viability of the design with respect to the reference CANDU kinf value.

The goal is to change the amount of 19.5 wt% U-235 enriched fuel used in the new

designs while maintaining the overall fuel bundle geometry such that the overall bundle

reactivity is compatible with the reference CANDU 37-element lattice.

For Designs 1, 2 and 3, the initial estimates of the thickness (or radius) of the

enriched fuel region are made based upon a simple mass balance relation with-respect-

to the mass of U-235 in new designs and the standard CANDU bundle, while ignoring

bundle physics related parameters such as bundle geometry, neutron flux and parasitic

absorptions. These estimations are provided in Appendix C.

The initial estimates are the starting point for inputting new fuel specifications in

the unitcell lattice simulations. The initial estimates are later adjusted based on their

deviation from the data collected using the reference CANDU lattice specifications.

Following the first few data points collection for new designs, the later data points are

synthesized through linear interpolation.

Designs 4 and 5 are slight variations of Designs 2 and 3, respectively, where the

initial concept of using mixed fuel (enriched + depleted) in Ring 4 elements are extended

to all rings for maximum yields of Mo-99 per bundle. Hence, the initial estimates for the

thickness of the enriched part was inferred from the parent designs (Design 2 or 3), and

altered accordingly based on the DRAGON simulation results.

Page | 79

Due to the higher fuel density of uranium metal compared to ceramic UO2 fuel,

Designs 3 and 5 present a challenge for balancing the mass of fuel in the pins, and

avoiding any resulting dissimilarity in bundle ring power distribution. The solution is to

find an optimal cladding thickness that would preserve the compatibility of bundle

reactivity with the existing CANDU bundle. The proposed thickness of the cladding

(2.27 mm) was calculated through mass balancing of both U-235 and U-238 in Design 3

and Design 5 fuel elements with the reference CANDU bundle.

The addition of enriched uranium in the Mo-99 target pins causes an increase in

fission rate resulting in a higher reactivity in that specific region. To compensate for this

increase reactivity, the remainder of the fuel pin is composed of depleted uranium.

Multiple cases with varying thickness (or radius) of the enriched region are

simulated in DRAGON for each design until a thickness (or radius) is determined with Δρ

value within the acceptable range (±1 mk). These cases are summarized in Tables 4-4

through 4-8 for each design. It is apparent from DRAGON results that the lattice

reactivity is very sensitive to changes in the amount of enriched region for the fixed

enrichment of 19.5 wt% U-235. Small changes in the amount of the enriched fuel

produce relatively large increase in reactivity.

Figures 4-3 through 4-7 illustrate the unitcell eigenvalue (k-infinity) as a function

of the thickness (or radius) of the enriched fuel region for the same overall fuel pin radius

as the reference bundle and compared with the kinf of the CANDU lattice.

Page | 80

Design 1

Table 4-4: Design 1 Unitcell Lattice k-infinity for Variations in Radius of

Enriched Central Region

Case # Radius of Enriched Central Target Region (mm)

Fresh Fuel k-infinity

∆ρ (mk)

1 1.00 1.0874 -29.0 2 1.10 1.1158 -5.64 3 1.20 1.1418 14.7 4* 1.14 1.1263 2.72 5* 1.127 1.1229 0.0301

* Value obtained through linear interpolation of data points

Figure 4-3: Design 1 Unitcell Lattice k-infinity Versus Variations in Radius of

Enriched Central Target Region (Fuel Pin Radius 6.075 mm)

1.060

1.070

1.080

1.090

1.100

1.110

1.120

1.130

1.140

1.150

1 mm 1.1 mm 1.2 mm 1.14 mm 1.127 mm

K-I

NF

INIT

Y

Design 1 Reference CANDU

Page | 81

Design 2

Table 4-5: Design 2 Unitcell Lattice k-infinity for Variation in Thickness of

Enriched Annular Region

Case #

Thickness of Enriched Annular Target Region (mm)


∆ρ (mk)

1 0.0750 1.1067 -13.0 2 0.125 1.1961 54.5 3 0.0850 1.1267 3.07 4* 0.0830 1.1228 0.00793


Figure 4-4: Design 2 Unitcell Lattice k-infinity Versus Variations in Thickness

of Enriched Annular Target Region (Fuel Pin Radius 6.075 mm)

1.070

1.090

1.110

1.130

1.150

1.170

1.190

0.075 mm 0.125 mm 0.085 mm 0.083 mm

K-I

NF

INIT

Y


Page | 82

Design 3



Case #



∆ρ (mk)

1 0.0660 1.1357 10.1 2 0.0560 1.1090 -11.1 3* 0.0610 1.1227 -0.108




1.070

1.080

1.090

1.100

1.110

1.120

1.130

1.140

1.150

1.160

0.066 mm 0.056 mm 0.061 mm

K-I

NF

INIT

Y


Page | 83

Design 4



Case #



∆ρ (mk)

1 0.0830 1.1248 1.54 2 0.0820 1.1212 -1.29 3* 0.0825 1.1230 0.128




1.080

1.090

1.100

1.110

1.120

1.130

1.140

1.150

0.0830 mm 0.0820 mm 0.0825 mm

K-I

NF

INIT

Y


Page | 84

Design 5



Case #

Thickness of Enriched Annular Region (mm)


∆ρ (mk)

1 0.0610 1.1201 -2.15 2 0.0620 1.1246 1.37 3* 0.0616 1.1228 -0.0325


Figure 4-7: Design 5 Unitcell Lattice k-infinity Versus Variations in

Thickness of Enriched Annular Target Region (Fuel Pin Radius 4.266 mm)

1.080

1.090

1.100

1.110

1.120

1.130

1.140

1.150

0.061 mm 0.062 mm 0.0616 mm

K-I

NF

INIT

Y


Page | 85

4.2 Task 2 - Comparison of Linear Heat Rating

The new designs must maintain a maximum power rating licensing limit of

60 kW/m to allow for existing operational practices to be maintained. Furthermore, to

avoid dissimilar power peaking effects in the bundle, the new designs are intended to

mimic the existing 37-element bundle radial pin power profile.

As mentioned in the previous section, in order to deal with the increased

reactivity of the fuel pins containing enriched fuel region over NU fuel, the remaining fuel

material in the pins is made up of depleted uranium. Once the thickness (or radius) of

the enriched region in the pins for each design is selected in Task 1 whereby the total

change in reactivity is within ±1mk with-respect-to the reference CANDU lattice, the

designs with those particular thicknesses (or radii) are further investigated for

evaluation and calculation of the element linear heat rating and fuel centreline

temperatures.

Data structures for macroscopic fission cross-sections, condensed neutron

volume integrated flux, and 2-D volume etc., for each fuel region are obtained from

DRAGON and processed for calculating the bundle radial pin power profile. The linear

heat rating is calculated using the methodology described in Section 3.4.1. The DRAGON

data structures for linear heat rating calculations are provided in Tables 4-9 through

4-13.

Figures 4-8 through 4-12 show the power profiles of the new designs versus the

reference CANDU bundle located in the highest power channel in the core. The

Page | 86

corresponding bundle specific power is 45.34 W/g (based on 900 kW bundle power). All

the pin powers and linear ratings are scaled to this specific power.

As expected, the power profile is lowest in the centre and higher in the outer rings

due to the corresponding neutron flux at these locations. The radial pin power profiles

of Mo-99 target bundles are also very similar to the reference CANDU radial pin profile,

whereby all designs meets the linear heat rating licensing limit of 60 kW/m with

adequate margins.

The linear heat rating of the target elements in the modified fuel bundle designs

was a priori postulated to be higher than the reference CANDU bundle due to the

presence of enriched fuel region. The resemblances of the two curves is not quite

unexpected. Efforts to keep the fresh fuel kinf value for a partially enriched fuel bundles

as close as possible to the reference CANDU lattice resulted in a configuration which

consisted of a relatively large depleted uranium region and a relatively small enriched

fuel region in these pins. The volumetric ratio of enriched fuel to depleted fuel was

3.56% in Design 1, 2.79% in Design 2 and 4, and 2.92% in Design 3 and 5. Therefore, the

increase in the fission rate and thus power density in the enriched region is compensated

by the relative decrease of power density in the depleted region, resulting in an overall

linear heat rating profile similar to the reference CANDU fuel bundle.

Page | 87

Design 1

Table 4-9: Design 1 Data for Radial Pin Power Profile Calculations

Region


Integrated Flux, ̂

(n·cm/s)

2-D Volume, V2-D (cm2)




Rating , '

rQ

(kW/m) Ring 1 0.02232 0.3637 1.159 2.658E-16 10.83 37.14 Ring 2 0.02331 2.194 6.956 1.675E-15 65.00 39.01 Ring 3 0.02593 4.484 13.91 3.807E-15 130.0 44.33 Ring 4 (Enriched)

0.5216 0.2538 0.7182 4.299E-15 6.711 56.32

Ring 4 (Depleted)

0.01329 6.714 20.15 2.957E-15 188.2

Figure 4-8: Design 1 Bundle Radial Pin Power Profile

30

35

40

45

50

55

60

1 2 3 4

LIN

EA

R H

EA

T R

AT

ING

(K

W/M

)

FUEL BUNDLE RINGS


Page | 88

Design 2


Region


Integrated Flux, ̂

(n·cm/s)

2-D Volume, V2-D

(cm2)




Rating , '

rQ

(kW/m) Ring 1 0.02236 0.3658 1.159 2.678E-16 10.83 37.43 Ring 2 0.02327 2.211 6.956 1.685E-15 65.00 39.26 Ring 3 0.02586 4.513 13.91 3.821E-15 130.0 44.50 Ring 4 (Depleted)

0.01337 6.771 20.30 2.999E-15 189.7 56.11

Ring 4 (Enriched)

0.6737 0.1932 0.5664 4.227E-15 5.292


30

35

40

45

50

55

60

1 2 3 4

LIN

EA

R H

EA

T R

AT

ING

(K

W/M

)

FUEL BUNDLE RINGS


Page | 89

Design 3


Region


Integrated Flux, ̂

(n·cm/s)

2-D Volume, V2-D

(cm2)




Rating , '

rQ

(kW/m) Ring 1 0.02243 0.3673 1.159 2.698E-16 10.83 37.67 Ring 2 0.02335 2.224 6.956 1.700E-15 65.00 39.58 Ring 3 0.02616 4.530 13.91 3.880E-15 130.0 45.15 Ring 4 (Depleted)

0.02589 3.336 9.999 2.860E-15 189.4 55.56

Ring 4 (Enriched)

1.319 0.1004 0.2922 4.302E-15 5.537


30

35

40

45

50

55

60

1 2 3 4

LIN

EA

R H

EA

T R

AT

ING

(K

W/M

)

FUEL BUNDLE RINGS


Page | 90

Design 4


Region


Integrated Flux, ̂

(n·cm/s)

2-D Volume, V2-D

(cm2)




Rating , '

rQ

(kW/m) Ring 1 (Depleted)

0.009508 0.3546 1.128 55.87 10.54 37.70

Ring 1 (Enriched)

0.4927 0.009897 0.03128 1.549 0.2922

Ring 2 (Depleted)

0.009894 2.143 6.769 335.2 63.24 39.52

Ring 2 (Enriched)

0.5111 0.05991 0.1877 9.294 1.753

Ring 3 (Depleted)

0.01100 4.372 13.53 670.5 126.5 44.90

Ring 3 (Enriched)

0.5686 0.1225 0.3753 18.59 3.507

Ring 4 (Depleted)

0.013348 6.760 20.30 1005 189.7 55.74

Ring 4 (Enriched)

0.6729 0.1918 0.5630 27.88 5.260


30

35

40

45

50

55

60

1 2 3 4

LIN

EA

R H

EA

T R

AT

ING

(K

W/M

)

FUEL BUNDLE RINGS


Page | 91

Design 5


Region


Integrated Flux, ̂

(n·cm/s)

2-D Volume, V2-D

(cm2)




Rating , '

rQ

(kW/m) Ring 1 (Depleted)

0.01873 0.1755 0.5553 1.089E-16 10.52 38.52

Ring 1 (Enriched)

0.9797 0.005259 0.01639 1.673E-16 0.3106

Ring 2 (Depleted)

0.01945 1.057 3.332 6.811E-16 63.14 39.89

Ring 2 (Enriched)

0.9896 0.03221 0.09835 1.035E-15 1.863

Ring 3 (Depleted)

0.02155 2.1509 6.664 1.535E-15 126.2 45.17

Ring 3 (Enriched)

1.110 0.06521 0.1967 2.350E-15 3.727

Ring 4 (Depleted)

0.02581 3.315 9.996 2.833E-15 189.4 55.40

Ring 4 (Enriched)

1.316 0.1009 0.2951 4.315E-15 5.591


30

35

40

45

50

55

60

1 2 3 4

LIN

EA

R H

EA

T R

AT

ING

(K

W/M

)

FUEL BUNDLE RINGS


Page | 92

4.3 Task 3 - Comparison of Radial Temperature Profiles

Another important fuel design parameter is fuel centreline temperature. The

margins available for fuel melting under normal operations and various accidental

scenarios play an important role in fuel bundle design and design optimization. In

addition to fuel centreline temperature, the sheath temperature must also be kept below

the accepted limits established by the industry.

The melting point of UO2 and U-metal is 2865 °C and 1132 °C, respectively. For

the purpose of this work, it is assumed that no gap is developed between the fuel and

cladding due to the short irradiation period required to reach peak activities of Mo-99.

The variation in UO2 thermal conductivity as a function of fuel temperature is accounted

for by utilizing the IAEA UO2 thermal conductivity relation (Eq. 3.26). A second radial

temperature profile is produced using a conservative value of UO2 thermal conductivity

(0.024 W/cm∙k). The thermal conductivity of U-metal is 0.43 W/cm∙k.

The fuel temperature calculations are greatly dependent on the heat conduction

coefficient used. The intent of the analysis is to compare the radial temperature profiles

of the highest power element of the new designs to the reference CANDU bundle profile,

and to ensure that the fuel centreline melting criterion is satisfied. This is illustrated in

Figures 4-13 through 4-20.

Thermophysical properties of coolant obtained using NIST REFPROP are

provided in Table 4-14. Table 4-15 lists the CANDU specific important heat transfer

parameters required for radial temperature profile calculations.

Page | 93

Additional data on materials and lattice specifications for 37-element CANDU fuel

bundle are provided in Appendix B. A numerical integration methodology as described

in Section 3.4.2 is used to calculate the radial temperature profiles for all designs, and

summarized in Tables 4-16 through 4-25.

Using the numerical integration method, the centreline temperatures of the

Mo-99 target bundles are calculated and found to satisfy the centreline melting criterion

for all new designs with the exception of Design 1 (See Figures 4-13 and 4-14). This is

not unexpected. The UO2 centreline temperature has always been a concern for the

industry due to relatively low thermal conductivity of ceramic UO2. In the case of

Design 1, the 19.5 wt% U-235 enriched fuel region is located in the centre of the Mo-99

target pins. The enriched fuel region with its high power density is encapsulated by a

relatively large volume of depleted fuel with much lower power density (see Table 4-16).

The power density of the enriched region is approximately 43 times higher than the

depleted region. Hence, the majority of the power production takes place towards the

centre of the pin resulting in significantly high centreline temperatures.

For the remaining designs, the fuel centreline temperature is lower than the

reference fuel. This is due to the relatively small thickness of the enriched fuel region on

the periphery of the fuel elements.

For the same steady-state channel conditions used in this analysis, the ΔT across

the thicker cladding used in Designs 3 and 5 is much higher compared to the reference

bundle. This is compensated by the much higher thermal conductivity of the U-metal

Page | 94

(0.43 W/cm∙K) as compared to ceramic UO2. The resulting centreline temperatures

provide adequate margins to centreline melting.

Table 4-14: Thermophysical Properties of Coolant Using NIST REFPROP [46]

Temperature (°C)

Pressure (MPa)

Density (kg/m3)

Viscosity (Pa∙s) x 10-6

Prandtl Number

Channel Inlet

266 11.05 781.9 101.0 0.8215

Channel Outlet

312 10.29 686.5 81.41 0.9534

Table 4-15: CANDU Specific Important Heat Transfer Parameters [46]

Parameter Value

Average Coolant Velocity through the Channel, V 10 m/s Coolant Flow Area, A 0.003507 m2

Wetted Perimeter, Pw 1.847 m Hydraulic Diameter, Dh 0.007595 m Reynolds Number, Re 305661 Nusselt Number, Nu 536.0 Coolant Heat Transfer Coefficient, HC 4.006 W/cm2∙K Thermal Conductivity of Clad, kC 0.107 W/cm∙K Thermal Conductivity of UO2 0.024 W/cm∙K Thermal Conductivity of U-Metal 0.430 W/cm∙K

Page | 95

Design 1

Table 4-16: Design 1 Normalized Heat Density and Linear Power Rating in

Ring 4 Sub-regions


Enrichment (wt %)

Thickness (cm)

Pin Radii (cm)



Enriched1 19.50 0.03757 0.03757 7797 34.57 Enriched2 19.50 0.03757 0.07513 8057 141.7 Enriched3 19.50 0.03757 0.1127 8659 333.7 Depleted1 0.2000 0.04948 0.1622 194.8 342.0 Depleted2 0.2000 0.04948 0.2117 198.5 353.5 Depleted3 0.2000 0.04948 0.2611 200.6 368.3 Depleted4 0.2000 0.04948 0.3106 202.0 386.2 Depleted5 0.2000 0.04948 0.3601 203.3 407.4 Depleted6 0.2000 0.04948 0.4096 204.4 431.9 Depleted7 0.2000 0.04948 0.4591 205.5 459.7 Depleted8 0.2000 0.04948 0.5085 206.7 490.8 Depleted9 0.2000 0.04948 0.5580 207.9 525.2 Depleted10 0.2000 0.04948 0.6075 209.5 563.2

Page | 96

Table 4-17: Design 1 Numerical Integration Results Using IAEA and Constant

Thermal Conductivity for UO2

Node Temperature Using IAEA Thermal Conductivity (°C)

Temperature Using Constant Thermal Conductivity (°C)

Fuel Centreline 4486 5862 Node1 (Enriched1) 4428 5748 Node2 (Enriched2) 4247 5398 Node3 (Enriched3) 3921 4794 Node4 (Depleted1) 3425 3962 Node5 (Depleted2) 2996 3342 Node6 (Depleted3) 2580 2837 Node7 (Depleted4) 2149 2401 Node8 (Depleted5) 1705 2012 Node9 (Depleted6) 1305 1653 Node10 (Depleted7) 987.3 1316 Node11 (Depleted8) 738.1 993.5 Node12 (Depleted9) 538.4 680.7 Node13 (Depleted10) 374.1 374.1 Node14 (Sheath) 310.0 310.0 Node15 (Coolant) 275.8 275.8

Page | 97

Figure 4-13: Design 1 Radial Temperature Profile of Ring 4 Pin Using IAEA


0

500

1000

1500

2000

2500

3000

3500

4000

4500

5000

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8

Te

mp

era

ture

(°C

)


Design 1

ReferenceCANDU

EnrichedRegion

DepletedRegion

Fuel Centreline

UO2 Melting Point

Page | 98

Figure 4-14: Design 1 Radial Temperature Profile of Ring 4 Pin Using a

Constant Thermal Conductivity for UO2 (0.024 W/cm∙K)

0

1000

2000

3000

4000

5000

6000

7000

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8

Te

mp

era

ture

(°C

)


Design 1

ReferenceCANDU

EnrichedRegion

DepletedRegion

Fuel Centreline

UO2 Melting Point

Page | 99

Design 2


Ring 4 Sub-regions


Enrichment (wt %)

Thickness (cm)

Pin Radii (cm)



Depleted1 0.2000 0.05992 0.05992 204.5 2.306 Depleted2 0.2000 0.05992 0.1198 204.6 9.233 Depleted3 0.2000 0.05992 0.1798 204.7 20.78 Depleted4 0.2000 0.05992 0.2397 205.1 36.97 Depleted5 0.2000 0.05992 0.2996 205.3 57.82 Depleted6 0.2000 0.05992 0.3595 205.7 83.35 Depleted7 0.2000 0.05992 0.4194 206.2 113.6 Depleted8 0.2000 0.05992 0.4794 206.7 148.5 Depleted9 0.2000 0.05992 0.5393 207.3 188.3 Depleted10 0.2000 0.05992 0.5992 207.8 232.8 Enriched1 19.50 0.002767 0.6020 10444 341.9 Enriched2 19.50 0.002767 0.6047 10384 450.8 Enriched3 19.50 0.002767 0.6075 10466 561.1

Page | 100





Fuel Centreline 874.8 1178 Node1 (Depleted1) 868.8 1171 Node2 (Depleted2) 851.1 1148 Node3 (Depleted3) 822.0 1109 Node4 (Depleted4) 782.5 1056 Node5 (Depleted5) 733.4 987.1 Node6 (Depleted6) 676.1 902.7 Node7 (Depleted7) 611.9 802.9 Node8 (Depleted8) 542.1 687.5 Node9 (Depleted9) 468.3 556.5 Node10 (Depleted10) 391.7 409.9 Node11 (Enriched1) 387.3 401.1 Node12 (Enriched2) 381.3 389.1 Node13 (Enriched3) 373.8 373.8 Node14 (Sheath) 309.9 309.9 Node15 (Coolant) 275.8 275.8

Page | 101



0

500

1000

1500

2000

2500

3000

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8

Te

mp

era

ture

(°C

)


Design 2

ReferenceCANDU

EnrichedRegion

DepletedRegion

Fuel Centreline

UO2 Melting Point

320

370

420

470

0.590 0.600 0.610 0.620

Tem

per

atu

re (

°C)


Page | 102



0

500

1000

1500

2000

2500

3000

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8

Te

mp

era

ture

(°C

)


Design 2

ReferenceCANDU

EnrichedRegion

DepletedRegion

Fuel Centreline

UO2 Melting Point

320

370

420

470

0.59 0.60 0.61 0.62

Te

mp

era

ture

(°C

)


Page | 103

Design 3


Ring 4 Sub-regions


Enrichment (wt %)

Thickness (cm)

Pin Radii (cm)




Page | 104

Table 4-21: Design 3 Numerical Integration Results Using Constant Thermal

Conductivity for U-Metal

Node Temperature Using Constant Thermal Conductivity (°C)

Fuel Centreline 793.0 Node1 (Depleted1) 792.6 Node2 (Depleted2) 791.4 Node3 (Depleted3) 789.3 Node4 (Depleted4) 786.5 Node5 (Depleted5) 782.8 Node6 (Depleted6) 778.4 Node7 (Depleted7) 773.1 Node8 (Depleted8) 766.9 Node9 (Depleted9) 760.0 Node10 (Depleted10) 752.2 Node11 (Enriched1) 751.7 Node12 (Enriched2) 751.0 Node13 (Enriched3) 750.1 Node14 (Sheath) 309.6 Node15 (Coolant) 275.8

Page | 105


Constant Thermal Conductivity for U-Metal (0.43 W/cm∙K)

0

500

1000

1500

2000

2500

3000

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8

Te

mp

era

ture

(°C

)


Design 3

ReferenceCANDU

EnrichedRegion

DepletedRegion

Fuel Centreline

UO2 Melting Point

U-Metal Melting Point

720

740

760

780

0.41 0.42 0.43 0.44

Tem

per

atu

re (

°C)


Page | 106

Design 4


Ring 4 Sub-regions


Enrichment (wt %)

Thickness (cm)

Pin Radii (cm)




Page | 107





Fuel Centreline 871.3 1175 Node1 (Depleted1) 865.4 1167 Node2 (Depleted2) 847.7 1144 Node3 (Depleted3) 818.9 1106 Node4 (Depleted4) 779.6 1052 Node5 (Depleted5) 730.8 984.1 Node6 (Depleted6) 673.8 900.0 Node7 (Depleted7) 609.9 800.4 Node8 (Depleted8) 540.6 685.5 Node9 (Depleted9) 467.1 554.9 Node10 (Depleted10) 390.9 408.8 Node11 (Enriched1) 386.5 400.2 Node12 (Enriched2) 380.6 388.2 Node13 (Enriched3) 373.1 373.1 Node14 (Sheath) 309.7 309.7 Node15 (Coolant) 275.8 275.8

Page | 108



0

500

1000

1500

2000

2500

3000

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8

Te

mp

era

ture

(°C

)


Design 4

ReferenceCANDU

EnrichedRegion

DepletedRegion

Fuel Centreline

Page | 109



0

500

1000

1500

2000

2500

3000

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8

Te

mp

era

ture

(°C

)


Design 4

ReferenceCANDU

EnrichedRegion

DepletedRegion

Fuel Centreline

UO2 Melting Point

Page | 110

Design 5


Ring 4 Sub-regions


Enrichment (wt %)

Thickness (cm)

Pin Radii (cm)




Page | 111

Table 4-25: Design 5 Numerical Integration Results Using Constant Thermal

Conductivity for U-Metal

Node Temperature Using Constant Thermal Conductivity (°C)

Fuel Centreline 791.2 Node1 (Depleted1) 790.8 Node2 (Depleted2) 789.6 Node3 (Depleted3) 787.6 Node4 (Depleted4) 784.8 Node5 (Depleted5) 781.1 Node6 (Depleted6) 776.7 Node7 (Depleted7) 771.5 Node8 (Depleted8) 765.4 Node9 (Depleted9) 758.5 Node10 (Depleted10) 750.8 Node11 (Enriched1) 750.3 Node12 (Enriched2) 749.6 Node13 (Enriched3) 748.7 Node14 (Sheath) 309.5 Node15 (Coolant) 275.8

Page | 112


Constant Thermal Conductivity for U-Metal (0.43 W/cm∙K)

0

500

1000

1500

2000

2500

3000

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8

Te

mp

era

ture

(°C

)


Design 5

Reference CANDU

EnrichedRegion

DepletedRegion

Fuel Centreline

UO2 Melting Point

U-Metal Melting Point

Page | 113

4.4 Mass and Activities of Molybdenum-99

The proposed designs utilize 19.5 wt% U-235 enriched fuel over several target

pins which can generate a large yield of Mo-99 per bundle. Tasks 1, 2 and 3 ensure that

high yields of Mo-99 can be done without significantly changing the bundle physics and

fuel safety characteristics.

The mass and activity of Mo-99 is calculated for each design by extension of the

DRAGON calculated fission rates as described in Section 3.4.3. The total Mo-99

production values are 2335, 2297, 2336, 4131 and 4268 six-day Curies for Design 1 to 5,

respectively. This yield corresponds to approximately 19% of the world weekly demand

for Designs 1, 2 and 3, and approximately 34% of the world weekly demand for Designs 4

and 5.

The activities of Mo-99 produced per bundle at EOB and six-day Curies are listed

in Tables 4-26 through 4-30. The required extraction rate to meet worldwide demands

of Mo-99 for each design is also specified. A production cycle of 6 bundles per week for

Designs 1, 2 and 3, or 3 bundles per week for Designs 4 and 5 would yield sufficient

quantities of Mo-99 to ensure stable supply of Mo-99 to satisfy the global demand.

Page | 114

Table 4-26: Design 1 Molybdenum-99 Production Results

Bundle Location

EOB Activity of Mo-99 (Ci)

Total EOB Activity of Mo-99

Total EOB Mass of Mo-99

Total Six-Day Curies of Mo-99

Weekly Extraction Rate Required to Meet Worldwide Demand

Ring 1 0 15006 Ci/Bundle

31.26 mg/Bundle

2335 Ci/Bundle

~6 Bundles/Week

Ring 2 0 Ring 3 0 Ring 4 15006


Bundle Location






Ring 1 0 14761 Ci/Bundle

30.74 mg/Bundle

2297 Ci/Bundle

~6 Bundles/Week



Bundle Location






Ring 1 0 15008.21 Ci/Bundle

31.26 mg/Bundle

2336 Ci/Bundle

~6 Bundles/Week


Page | 115


Bundle Location






Ring 1 551.8 26547 Ci/Bundle

55.29 mg/Bundle

4131 Ci/Bundle

~3 Bundles/Week

Ring 2 3470 Ring 3 7894 Ring 4 14629


Bundle Location






Ring 1 583.0 27423 Ci/Bundle

57.11 mg/Bundle

4268 Ci/Bundle

~3 Bundles/Week

Ring 2 3606 Ring 3 8192 Ring 4 15040

Page | 116

Although the centre and middle rings provide higher margins to heating limits as

depicted in Figures 4-8 through 4-12, the 4th ring provides a higher flux exposure which

is a key factor for the production of Mo-99. This aspect of core neutronics is utilized in

Designs 1, 2 and 3. Furthermore, by only using the outer most rings for Mo-99

production, the post processing procedure becomes simpler since only those pins will

need to be demounted for processing, and can easily be replaced with fresh Mo-99

targets for another production run. Ideally the re-use of the remaining pins should be

limited to a few cycles since excessive irradiation of the NU will produce Pu-239 through

neutron absorption. Build-up of small amounts of Pu-239 will also occur in the depleted

fuel inside the Mo-99 producing pins. However, due to the 20-day irradiation period and

the presence of only a small number Mo-99 producing bundles in the core, the effect of

Pu-239 build-up on the overall reactivity is expected to be small.

The production of Mo-99 must be balanced against the bundle power and linear

heating limits. Assessment of Designs 1, 2 and 3 showed that the bundle reactivity and

distribution of power resembles greatly with the standard CANDU bundle at the pin

level. This resemblance made it possible to extend the design of the 4th ring elements to

other rings for maximum Mo-99 yield. Tables 4-26 through 4-30 list the total six-day

Curies of Mo-99 produced for each design. As calculated in these tables, Designs 4 and 5

distribute the total power of the bundle over increased number of target elements, 37

pins in total, and therefore provide a much higher yield of Mo-99 compared to Designs 1,

2 and 3.

Page | 117

Chapter 5

Concluding Remarks and Future Work

This thesis explores the possibility of using modified 37-element fuel bundles for

the production of Mo-99 in existing CANDU reactors without affecting the core

properties or day-to-day operations of the plant. The tasks performed in Sections 4.1,

4.2 and 4.3 are used to evaluate the viability of the designs with respect to the existing

safety margins as well as to demonstrate the similarity of the proposed target bundles

to the standard 37 element bundle behavior.

As mentioned above, various designs with different amounts of 19.5 wt% U-235

enriched fuel were simulated using the DRAGON lattice code and the fresh fuel kinf values

were recorded for each case. The use of 19.5 wt% U-235 enriched fuel as Mo-99 targets

in different rings of a 37-element fuel bundle result in increased fission rates and hence

power rating of the enriched region is increased. In order to have a high confidence in

the available margins to the maximum linear power ratings and centreline temperature

limits, the bundles are assumed be irradiated in the highest power channel of the CANDU

reactor core corresponding to the maximum instantaneous power of 900 kWth.

In conclusion, the design of modified fuel bundles for Mo-99 production met the

criteria specified in Section 3.1 of this thesis. Although the impacts of the enriched target

pins on safety analysis were not directly examined in this work, the similarity in the

unitcell eigenvalue (kinf) and the pin linear heat rating with the reference CANDU bundle

suggest that the resultant impact on accident analysis should be small. However, direct

Page | 118

confirmation may still be required before the new design can be implemented in

operating reactors.

Over the last few years, there have been a number of supply shortages of Mo-99.

The situation is to deteriorate further with the expected permanent shutdown of the

Canadian NRU reactor in 2016. The CANDU power reactors can play a large role in

ensuring a stable supply of Mo-99. The feasibility of producing large and continuous

quantities of Mo-99 in the CANDU power reactors has been demonstrated in this thesis.

The continuous and reliable supply of medical isotopes will ensure the availability of life-

saving medical procedures for millions of patients around the globe.

Recommendations for Future Work:

During the course of this research, some areas were identified as requiring future

study if the production of medical isotopes using one of the proposed bundle designs in

CANDU reactor was actually begun. These areas of potential research include:

Fuel pellet manufacturability: Designs 2, 3, 4 and 5 proposed in this thesis have

very thin annular enriched fuel regions composed of either UO2 or U-metal on the

periphery of the fuel pellet. Although the new fuel manufacturing methods have

significantly advanced, the proposed thin layers may still challenge manufacturing

feasibility. The manufacturability of the posed fuel pellets should be examined.

Use of demountable fuel bundles: The study assumes the use of demountable

bundles for the proposed designs. Upon completion of the irradiation cycle, the target

pins will be demounted and sent for processing. Although, this is not critical to the

Page | 119

proposed study, it will provide ease with respect to fuel reprocessing and reuse. The

feasibility of demountable bundles to use in the operating CANDU reactors should be

examined.

Handling and processing of medical isotopes producing 37-element bundle: Upon

extraction from the reactor the pins with enriched uranium will be sent to the Mo-99

production facilities for further processing, whereas the remaining pins can either be

sent away to spent fuel storage or be mounted on another target fuel bundle and

reloaded into the reactor for another Mo-99 production cycle. Additional fuel handling

procedures would need to be developed to remove the Mo-99 target pins from the

demountable bundle and send them for processing. Handling and processing of newly

designed fuel bundles would require updating the existing infrastructure and is a subject

for future research.

Pilot production: The study assumes that the proposed bundle will be irradiated

in the highest power channel of the CANDU reactor. The 900 kWth bundle power is the

upper limit for a typical CANDU fuel bundle and the bundle power at the periphery

location can be half of this value [50]. Ideally the new bundles will be first irradiated in

a low-power channel on the periphery of the core to get more operational experience

before moving the bundles to the high-power central channels. Therefore, from an

operational standpoint, it is advisable to load the new proposed bundles in the periphery

channels first to gain operational experience before loading them into the high power

channels of the CANDU core.

Page | 120

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Appendices

Appendix A – Potential Producers of Reactor Produced

Molybdenum-99

Numerous projects have surfaced worldwide related to the building of new

reactors dedicated to the Mo-99 production, including production of Mo-99 from an

existing facility that was not originally built for that purpose. In some cases, these

initiatives are expected to play a big role in the production of medical isotopes in the

future, while in other cases, the projects are either cancelled or lobbying for

governments to provide support for further research to be carried out.

The following is a list of potential producers of reactor-based Mo-99 that have the

technological resources to play a large role in the production of medical isotopes in the

future.

A.1 Multipurpose Applied Physics Lattice Experiment (MAPLE)

Designed by Atomic Energy of Canada Limited (AECL), two reactors, MAPLE I and

II were planned and constructed to supply the world’s demand for Mo-99. The MAPLE

reactors were designed to generate 10 MW of heat for the sole purpose of medical

isotope production [51]. Each reactor was capable of producing enough medical

isotopes to supply 100% of the worldwide demand (~12000 six-day Curies per week).

The MAPLE reactors were planned to use LEU fuel and HEU targets.

Page | 136

Designed and built to replace the 54 year old NRU reactor in Canada, the project

was estimated to cost $140 million but ended up costing much more due technical design

flaws which were discovered during commissioning. It was discovered that these

reactors have a positive power coefficient of reactivity. The reactors were designed to

have a small negative power coefficient of reactivity, estimated at -0.12 mk/MW, but it

was measured to be +0.28 mk/MW.

The power coefficient of reactivity describes the number of neutrons available

per generation per unit change in power. A positive value indicates that there are more

neutrons available in each subsequent generation as the reactor power is increased. This

meant that the reactor inherently tried to increase the power during power increases, a

positive feedback. The original safety systems of the reactor were inadequate and design

changes were implemented to accommodate the positive power coefficient of reactivity.

Investigations by AECL and other contracted service providers were unable to explain

the positive power coefficient, and in 2008, AECL and the Government of Canada

announced that the project would be discontinued [52].

There has been much debate among experts on the issue of reviving the MAPLE

reactors. A number of scientists believe that the reactor could have operated safety with

a positive power coefficient of reactivity, after the design changes to take it into account

were implemented. However, another roadblock hindering the MAPLE project was that

the fuel processing facility was designed to accommodate HEU targets. Conversion to

LEU targets would have required a significant redesign of the processing facility with

Page | 137

larger waste tanks, etc. Calls to complete commissioning of the MAPLE project have not

received much attention.

It is expected that a general re-revisit of the positive power reactivity coefficient

problem may take up to 6 years to rectify if possible. If it is later decided that redesigning

and replacing the entire core of MAPLE reactors is the only way to solve the positive

power reactivity coefficient problem, those efforts may take additional 7 to 10 years and

may cost tens of millions of dollars [53].

A.2 Aqueous Homogeneous Reactor (AHR)

Babcock & Wilcox (B&W) have proposed to build an aqueous homogeneous

reactor (AHR), also known as a solution reactor, which will utilize an LEU fuel salt

dissolved in acid and water. In this reactor, the uranium in the solution is both the

reactor fuel and the target material. This system would eliminate the need to

manufacture or encapsulate a target, and un-fissioned uranium can be recycled back into

the reactor instead of being removed as waste. Various solution forms of fuel are

possible, such as: uranyl nitrate (UO2(NO3)2), uranyl sulfate (UO2SO4), or uranyl fluoride

(UO2F2) [54]. The extraction process itself would not differ remarkably from the existing

adsorption process of using an alumina column that is currently used by processing

facilities.

B&W has estimated that an AHR reactor could generate approximately 18% of

the North American Mo-99 demand (1100 six-day Curies per week) while assuming a 26

Page | 138

hour processing time. The proposal suggests building three to four 200 kW units for

higher Mo-99 yields. They have also projected a total waste volume of less than 5 m3

each year consisting generally of alumina column wastes [54].

B&W assumes that five years will be required from conceptual design (which has

begun) to commercial operation for their newly designed AHR. However, the plan is still

lacking concrete budget projections derived from actual experience from building

solution reactors. B&W’s own initial estimate states that less than $100 million will be

required for research and development of the AHR [55].

Being first of its kind in terms of design and scale, another hurdle to the

construction of AHR is lack of regulatory framework. Current nuclear reactor

regulations do not specifically address AHR, so licensing under current regulations may

require clarifications, or new regulations may be needed to appropriately address the

safety of the solution reactors [55].

A.3 Open Pool Australian Light-Water (OPAL) Reactor

The OPAL reactor started its operation in late 2006. This multipurpose pool type

reactor was built by INVAP for the Australian Nuclear Science and Technology

Organization (ANSTO). The purpose was to replace the aging High Flux Australian

Reactor.

Page | 139

OPAL is a 20 MW reactor with peak thermal neutron flux: 3x1014 n·cm-2·s-1 in the

reflector vessel. The core of OPAL consists of compact 16 uranium oxide fuel assemblies

interspersed with control rods. The fuel assemblies are cooled by demineralised light

water, while the core is surrounded by heavy water moderator in a zirconium alloy

reflector vessel at the bottom of a 13-metre-deep open pool of light water [56].

OPAL uses LEU fuel with around 20% U-235 enrichment by weight [56]. This

gives OPAL a distinct advantage over other research reactors around the world in terms

of nuclear security and safeguards. OPAL has the capacity to operate 340 days a year

which is a significant increase over the operating levels typically achieved by comparable

research facilities.

The ANSTO facility includes an onsite processing facility which produces many

important radionuclides, such as Mo-99, extracted from LEU uranium-aluminium alloy

targets. The processing facility was also build by INVAP with a budget of $200 Million

[57]. The facility supplies medical isotopes to meet 100% of Australia’s demand, as well

as the demands of a number of overseas countries, especially in the Asia-Pacific region.

The reactor has been producing medical isotopes since May 2008, and is expected to

produce medical isotopes for the next 40 years.

The reactor is currently producing Mo-99 one manufacturing run per week. If

there is a need to produce for the wider market, such as North America, the

manufacturing process can be run multiple times per week with multiple targets being

irradiated [55]. However, before the operators of OPAL decide on expanding their

manufacturing runs, appropriate approvals by the government drug oversight agencies

Page | 140

will be required in order to use Mo-99 made by ANSTO. In July 2006, the US Food and

Drug Administration (FDA) and Health Canada authorized Lantheus Medical Imaging to

use Mo-99 produced by the OPAL reactor to make Tc-99m for diagnostic imaging for

North American health care facilities ([53], [58]).

Beside some political and long-distance transportation related issues, the OPAL

reactor was intended to only provide medical isotopes for the local Australian and New

Zealand markets, and is only capable of providing a small fraction of the worldwide

demand of Mo-99 ([53], [55]).

A.4 Jules Horowitz Reactor (RJH)

Jules Horowitz Reactor is a 100 MW material test reactor located at the Cadarache

site in southern France. Built by the French Commissariat à l'énergie atomique (Atomic

Energy Commission, CEA), the reactor is expected to replace the 70 MW OSIRIS reactor

when the RJH becomes operational. The reactor is projected to be operational in the

second quarter of 2016, with an estimated budget of 500 million Euros and with a

lifetime of 50 years ([59], [60]).

The RJH reactor core is dismountable type. Reactor fuel is inserted in the form of

cylinders of six concentric curved fuel plates hot-rolled from uranium alloys. 46

cylinders can fit in the one-piece aluminum alloy core rack at a time. The fuel is enriched

to 25% U-235 by weight, amounting to a total mass of 21 kg U-235. The dismountable

reactor core is partly surrounded by a beryllium reflector. The reactor core between

Page | 141

coolant inlet and outlet can be replaced to accommodate scope of future research

initiatives [60].

Although the facility and the related infrastructure was initially conceptualized

for scientific studies dealing with material and fuel behaviour under irradiation, a recent

focus on the production of medical isotopes as part of the European strategy to ensure

reliable supply for use in nuclear medicine has been gaining some momentum ([61],

[62]). The plan is to coordinate the production with the isotopes processing facilities in

the Netherlands to provide a back-up supply for medical isotopes for Europe.

RJH will have a thermal neutron flux of 5x1014 n·cm-2·s-1 in the reflector region

[63]. The reactor is expected to be operational for 220 days a year and can have flexible

Mo-99 production cycles. RJH is expected to produce 25% (1200 six-day Curies per

week) of the European base needs, with the possibility to increase the production up to

50% of European needs (2400 six-day Curies per week) if required. This can be achieved

by irradiating 32 to 48 LEU targets in the core and in the reflectors in both fixed and

movable positions [61].

Design studies for Mo-99 irradiation systems based on LEU targets has been

completed, and the irradiation tests of LEU targets are expected to take place in the

second quarter of 2018. The reactor is expected to begin its normal level of Mo-99

production in 2019 [62].

The RJH has the capacity to increase the number of target irradiation in order to

significantly contribute to the Mo-99 world market. However, the reactor is built mainly

Page | 142

for material testing purposes and is not likely to limit its activity to the production of

Mo-99 [64]. The project is expected to supply a fraction of the European medical isotope

need and is unlikely to play a major role in the North American medical isotope market.

A.5 China Advanced Research Reactor (CARR)

The China Advanced Research Reactor began its construction on August 26, 2002

and on March 1st, 2012 it successfully achieved full power operations ([65], [66]). The

reactor is located at the China Institute of Atomic Energy Fangshan, Beijing site. The

sophisticated light-water tank type reactor with a heavy water reflector is one of the

most advanced research reactors in the world, producing 60 MW at full power [66]. The

core of CARR is immersed in a 15-meter pool of water. CARR has maximum thermal

neutron flux of 1×1015 n·cm-2·s-1 at the centre of the reactor core, and about

8×1014 n·cm-2·s-1 at the heavy water reflector. CARR has nine tangent horizontal neutron

beam channels available for experiments and 25 vertical irradiation channels [67].

The overall design objective of CARR is to enhance China’s research capabilities

in fields such as nuclear physics and chemistry, neutron scattering experiments, testing

of reactor materials and nuclear fuels and neutron activation analysis. The reactor is

also being used to irradiate targets for the production of radioactive isotopes [65].

Mo-99 produced in CARR is sent to CIRA member companies for processing and

to produce Tc-99m generators. The CARR production capacity per week is about 250

six-day Curies of Mo-99. The demand activities of Mo-99 in China are 200 six-day Curies

Page | 143

per week. Therefore, CARR is expected to supply 100% of the China’s and neighbouring

countries medical isotope needs and is unlikely to play a major role in the North

American medical isotope market.

A.6 University of Missouri Research Reactor Center (MURR)

Built in 1966, the University of Missouri Research Reactor is a 10 MW tank type

reactor. MURR is the largest university research reactor in North America. It is a flux-

trap reactor with high intensity thermal and fast neutron flux which can be as great as

5x1014 n·cm-2·s-1 [68]. The reactor runs 24 hours a day, 7 days a week, with a scheduled

12 hour maintenance shutdown on Mondays. In 2006, the reactor was relicensed to

operate for another 20 years. The MURR is attempting to establish a domestic

production source of Mo-99. In 2007, the facility produced 41 different isotopes, and

more than 1000 shipments were exported to 14 different countries [69].

MURR has indicated its intention to commercially produce Mo-99 and is soliciting

funding for studying the design and construction of the processing facilities for post

irradiation extraction of medical isotopes [70]. Overall, the objective is to develop the

capability to produce Mo-99 from LEU targets, with little change to the current reactor

layout.

MURR has completed a conceptual design of a processing facility which will allow

the safe processing of the LEU targets. The facility is to be located adjacent to MURR to

permit safe transfer of targets to the facility after irradiation. Production objectives are

Page | 144

to produce 50% of current U.S. weekly demand (3000 six-day Curies per week) [71].

MURR could also help to fill gaps for up to about 75% of the US demand when other

reactors are on planned shutdowns. This can be achieved through short-term shifts in

production schedules [70]. It is estimated that it will take 3-4 years for MURR to build

the appropriate processing facilities to handle the anticipated volumes of Mo-99 [53].

The current status of the planned upgrades is unknown.

In March 2011, MURR entered into a contract with NorthStar to build an electron

linear accelerator to produce low specific activity Mo-99. The goal is to ramp up the

production to 3000 six-day Curies per week, but have only been able to produce just over

1100 six-day Curies ([72], [73]). Although the separation chemistry of the accelerator

produced low specific activity of Mo-99 is very well understood, the generated doses do

not fit in the current distribution stream. The low specific activity of Mo-99 requires new

generating systems to utilize the products and generate Tc-99m in activity

concentrations typically needed in nuclear pharmacies [73].

A.7 Sandia National Laboratories (SNL)

Home to the Annular Core Research Reactor (ACRR), the Sandia National

Laboratories are located on Kirtland Air Force Base in Albuquerque, New Mexico and are

managed and operated by the Sandia Corporation, which is a wholly owned subsidiary

of Lockheed Martin Corporation. Since the late 1990s, SNL has been seeking for licensing

approval to produce Mo-99. The initial preparatory work included reproduction and

Page | 145

verification of the Cintichem Mo-99 production process, development of production

hardware, modification of the hot cell facility, reconfiguration of the ACRR and

preparation of documentation to meet FDA and US-NRC approvals [74].

The ACRR is an open tank-type reactor filled with about 10 meters of water which

provides both core cooling and radiation shielding. Heat is transferred to the open water

pool through natural convection. The pool itself is cooled by an external heat exchanger.

The core of ACRR consists of an annular array of UO2-BeO fuel elements. The height of

the active fuel part is 52 centimetres. For target irradiation, the dry steel-lined control

cavity can be removed from the centre of the core to provide a water-filled region. Two

options are available for target irradiation, one with a maximum of 19 targets and the

other with a maximum of 37 targets at a time. With targets inserted, only 180 or 130

conventional fuel assemblies, depending on the option selected, would be required to

operate the reactor [75]. ACRR has flexibility to operate at 30000 MW in pulse mode or

4 MW in a steady state mode [55].

The SNL site which includes ACRR, hot cell facility, and other associated facilities

are a promising option for the isotope production program. The characteristics of ACRR

are fully compatible with radioisotope production. It is capable of being dedicated to

continuous isotope production, which is necessary to meet the demands for Mo-99. The

connected hot cell facility was modified in the 1990s for Mo-99 production. The ACRR

site is in proximity to excellent air transportation facilities which are useful for

radioisotope shipments [76].

Page | 146

The SNL facility has the capability to serve as a backup supply to provide up to

30 % of the U.S. market demand (~1800 six-day Curies per week) under normal

circumstances. It also has the capability to produce 100% of US Mo-99 requirement

(~6000 six-day Curies) should the need arise [75]. The problem however, is that the

reactor is currently being used by the defense programs group at the US-DOE. In order

to free the ACRR for production of medical radionuclides, it would require DOE to

reassign the mission of the ACRR from defense program uses, and make significant

investments in the reactor and the processing facilities [55].

It is estimated that it would require $10 to $50 million to convert the ACRR into

a steady state reactor and make necessary upgrades to the existing hot cell facility in

order to process Mo-99 using LEU targets [55]. Furthermore, the current fuel is 35%

enriched U-235, which is considered HEU. However, it is possible to convert the ACRR

to use LEU fuel for the production of Mo-99.

The proposed Mo-99 project regarding changing the mission of the ACRR for

Mo-99 production has been dropped by the DOE in favour of a new LEU reactor for the

production of Mo-99 at SNL. In the more recent publication coming out of SNL, the

facility is proposing to build a new reactor for the sole purpose of producing medical

isotopes alongside a hot cell facility ([77], [78] [79], [80]). The intention is to construct

a new reactor which is capable of producing 100% of the U.S demand (~6000 six-day

Curies per week). The reactor concept is unique in that the fuel for the reactor and the

targets for the Mo-99 production are the same. In other words, no driver core is required

for irradiation of Mo-99 targets. The fuel pins will be irradiated on a 7 to 21 day cycle.

Page | 147

The reactor will use LEU uranium oxide fuel enriched to 20 wt% of U-235. The fuel pins

will be approximately 1 cm in diameter and 30 to 40 cm in height, encapsulated in a

zirconium alloy cladding. An estimated 90 to 150 fuel pins will be arranged in the core

located inside a water pool about 9 meters deep [77].

The reactor power level will be between approximately 2 MW. Since the

production of 6000 six-day Curies per week requires at least 1.1 MW of continuous

target fission power, assuming two post-irradiation days for processing and shipping

[77], the proposed reactor is capable of meeting 100% of North American demand. The

proposed facility will also have an adjacent hot cell facility which would be connected to

the reactor pool by a water channel to facilitate remote transfer of irradiated targets for

processing. Processing will be done using oxide dissolution and separation processes. A

provisional patent for the SNL medical isotope reactor concept was filed on

September 11, 2009 [77].

Although, there are no impediments to prevent the SNL medical isotope reactor

concept along with its hot cell facility, from being designed, fabricated, and licensed

today, the only question is who will be funding this project. A report by US Society of

Nuclear Medicine (SNM) task group suggests that SNL should explore the possibility of

partnerships with National Nuclear Security Administration and DOE [55].

Page | 148

A.8 Other Potential Suppliers of Reactor Produced Molybdenum-99

Following the major disruptions in the supply of medical isotopes in 2007, the

Canadian government has initiated OECD12 meetings on radioisotope availability in Paris

and again in Toronto in 2009. Also in efforts to expand Canada's isotope production

capabilities, in May 2009 the provincial and federal governments announced a $22-

million fund for the McMaster University research reactor to upgrade the research

reactor’s infrastructure. However, the McMaster Nuclear Reactor (MNR) is a 50-year-

old reactor and will use HEU targets [53]. According to a report by SNM, even with 24/7

operation, it is estimated that MNR can only supply about 30% of North American Mo-99

demands [55]. The MNR is likely to assume the role of a backup supplier rather than a

primary producer of Mo-99 for North America.

In the United States, a proposal to produce Mo-99 by using LEU solution targets

at the University of California at Davis in the McClellan Nuclear Radiation Center reactor

could also reach the commercial production stage in 3-4 years [53].

Table A-1 provides a list of many potential reactor projects at various stages of

development. Some of these projects are very well advanced, while others are still in

proposal or a design phase seeking funding and/or regulatory approvals before actually

advancing to the construction phase.

12 The Organisation for Economic Co-operation and Development

Page | 149

Table A-1: Potential Future Suppliers of Reactor Produced Molybdenum-99

[13]

Reactor Facility

Target Expected Normal Capacity per Week (six-day Curies)

Estimated First Full Year of Production

Status

MNR (McMaster University, Canada)

LEU 1500 Project is seeking funding.

RIAR (Russian Federation)

HEU 800-2000 2015 The facility includes three reactors. Two of which will produce Mo-99, with the third being a back-up. Currently in Phase 2 of development.

INR, Pitesti (Romania)

LEU 900 2015 Proposal

FRM-II (Germany)

LEU 1950 2016 Pending LEU target design approval.

Coqui (United States)

LEU 7000 2017 Project is seeking funding.

BMR (Brazil) LEU 1000 2018 Project is in preliminary design stage.

RA-10 (Argentina)

LEU 200 2019 Project is in preliminary design stage.

PALLAS (Netherlands)

LEU 7300 2023 Project is in preliminary design stage. Project is also seeking funding.

SAFARI-II (South Africa)

LEU 3000 2026 Feasibility study for this project is underway.

Page | 150

A.9 New Approaches in Diagnostic Nuclear Medicine

Going into the future, there may be a number of accelerators and nuclear-based

technologies that could play a potential role in providing a stable supply of Mo-99. In

addition, there are a variety of new and emerging medical imaging techniques that could

possibly bypass the need for nuclear reactor based medical isotopes. Positron Emission

Tomography – Computed Tomography has already been successful in this regard.

However, there is no guarantee that any new technology will receive clinical acceptance

even though it may be commercially viable. Cost, ease of use, facilities for the production

and transportation may be some of the determinants [81].

Other means of producing Mo-99 include photo-fission of U-235, neutron

activation of Mo-98, and photo neutron interactions of Mo-100. Even before the 2007

crisis, TRIUMF based in Vancouver, Canada, proposed a plan to produce Mo-99 by

photon fission of U-238 by using photons from an electron accelerator [82]. This method

of production yields low quantities of Mo-99, which requires multiple units to supply the

world's demands.

Construction of new reactors or conversion of existing reactors as well as

possible use of accelerators to meet increasing demands has been the focus of the

medical isotope industry in recent years, including the use of alternate radionuclides and

medical imaging techniques ([2], [8], [11], [81], [83], [84]).

Page | 151

Appendix B – Materials, Lattice Specifications and Important

Parameters for 37-Element CANDU Fuel Bundle Calculations

Materials, Lattice Specifications and Other Important Parameters

Value

Initial Uranium Composition Natural UO2 U-235 U-238 Enriched UO2/U-metal U-235 U-238 Depleted UO2/U-metal U-235 U-238

Weight Percentage 0.711% 99.289% 19.50% 80.50% 0.200% 99.8%

UO2 fuel density U-metal fuel density

10.60 g/cm3 18.95 g/cm3

Fuel temperature 1155 K Element radius U-metal fuel element radius

0.6075 cm 0.4266 cm

Element length/bundle length 49.53 cm Number of fuel pins 37 Inner fuel ring radius (6) Middle fuel ring radius (12) Outer fuel ring radius (18)

1.4885 cm 2.8755 cm 4.3305 cm

Cladding material Cladding radius

Zircaloy 0.6540 cm

Pressure tube Inner radius Outer radius

5.1689 cm 5.6032

Calandria tube Inner radius Outer radius

Zircaloy 6.4478 cm 6.5875 cm

Coolant Atom purity Density Temperature

D2O 99.75 % 0.8360 g/cm3 549 K

Moderator Atom purity Density Temperature

D2O 99.91 % 1.083 g/cm3 346 K

Page | 152

Fuel channel square pitch 28.575 cm Approximate burnup 906.62 MWd/MgU Mo-99 fraction in fission products 6.1% Mo-99 decay constant (λ) 0.252 d-1

Mo-99 atomic weight 99.91 g/mol

Page | 153

Appendix C – Initial Estimates of Enriched Fuel Region Thickness and

Pin Radius for UO2 and U-Metal Fuel Designs Based on Uranium Mass

Balancing

Design 1

Fuel Pellet Radius = Rf = 0.6075cm

Fuel pin length = Lf = 49.5cm

Assumptions:

The isotopic composition of Uranium includes only U-235 and U-238 isotopes. Other uranium isotopes (U-234, U-236, etc.) are ignored.

Weight percentage of U-235 in natural uranium, rnat,U = 0.71% Weight percentage of U-235 in enriched uranium, renr,U = 19.5% Weight percentage of U-235 in depleted uranium, rdep,U = 0.2% Density of UO2 will remain the same irrespective of the enrichment, ρnat,U = ρdep,U = ρenr,U

= 10.6 g/cm3

Atomic weight of enriched Uranium, MU, is:

238

235

235

235

238

238

235

235 11

U

U

U

U

U

U

U

U

U M

r

M

r

M

r

M

r

M

Atomic weight of UO2 = 16*2 OU MM , therefore:

Mass of U-235 in CANDU-6 fuel pin is:

mnat,U-235 = Unat

OnatU

natUUnatff r

MM

MLR ,

16

,

2

*2

Mass of U-238 in CANDU-6 fuel pin is:

mnat,U-238 = )1(*2

,

16

,

2

Unat

OnatU

natUUnatff r

MM

MLR

In Design 1, the enriched UO2 fuel region is inscribed inside the depleted uranium annulus. The mass of U-235 in the enriched fuel pin is therefore:

m’enr,U-235 = Uenr

OenrU

enrUUenrfenrf r

MM

MLR ,

16

,

2

*2'

Page | 154

The mass of U-235 in depleted uranium annulus is:

m’dep,U-235 = Udep

OdepU

depU

Udepfenrff rMM

MLRR ,

16

,

22

*2)'(

To calculate the radius of the enriched fuel layer in Design 1, a simple mass balance relation can be formulated based on the mass of U-235 in the reference CANDU-6 and Design 1 fuel pins.

mnat,U-235 = m’enr, U-235 + m’dep, U-235

Udep

OdepU

depU

Udepfenrff

Uenr

OenrU

enrUUenrfenrfUnat

OnatU

natUUnatff

rMM

MLRR

rMM

MLRr

MM

MLR

,

16

,

22

,

16

,

2

,

16

,

2

*2)'(

*2'

*2

The above equation can be simplified to:

UdepUdepUdepenrffUenrUenrUenrenrfUnatUnatUnatf rRRrRrR ,,,

22

,,,

2

,,,

2 )'('

UnatUnatUnatfUdepUdepUdepf

UenrUenrUenrenrfUdepUdepUdepenrf

rRrR

rRrR

,,,

2

,,,

2

,,,

2

,,,

2 ''

(i)

where,

16

1

238

,

235

,

1

238

,

235

,

16

,

*21

1

*2

O

U

Unat

U

Unat

U

Unat

U

Unat

OnatU

natUUnat

MM

r

M

r

M

r

M

r

MM

M

16

1

238

,

235

,

1

238

,

235

,

16

,

*21

1

*2

O

U

Udep

U

Udep

U

Udep

U

Udep

OdepU

depU

Udep

MM

r

M

r

M

r

M

r

MM

M

16

1

238

,

235

,

1

238

,

235

,

16

,

*21

1

*2

O

U

Uenr

U

Uenr

U

Uenr

U

Uenr

OenrU

enrUUenr

MM

r

M

r

M

r

M

r

MM

M

Eq. (i) then becomes:

Page | 155

UenrUenrUenrUdepUdepUdep

UnatUnatUnatUdepUdepUdepf

enrfrr

rrRR

,,,,,,

,,,,,,

2

2

UenrUenrUenrUdepUdepUdep

UnatUnatUnatUdepUdepUdepf

enrfrr

rrRR

,,,,,,

,,,,,,

2

(ii)

Assuming the densities do not change for different enrichments:

UnatUdepUenr ,,,

Eq. (ii) can be simplified to:

UenrUenrUdepUdep

UnatUnatUdepUdepf

enrfrr

rrRR

,,,,

,,,,

2

where,

994.15*2051.238

%71.01

044.235

%71.0

051.238

%71.01

044.235

%71.0

*21

1

1

1

16

1

238

,

235

,

1

238

,

235

,

,

O

U

Unat

U

Unat

U

Unat

U

Unat

Unat

MM

r

M

r

M

r

M

r

8815.0, Unat

994.15*2051.238

%2.01

044.235

%2.0

051.238

%2.01

044.235

%2.0

*21

1

1

1

16

1

238

,

235

,

1

238

,

235

,

,

O

U

Udep

U

Udep

U

Udep

U

Udep

Udep

MM

r

M

r

M

r

M

r

8815.0, Udep

994.15*2051.238

%5.191

044.235

%5.19

051.238

%5.191

044.235

%5.19

*21

1

1

1

16

1

238

,

235

,

1

238

,

235

,

,

O

U

Uenr

U

Uenr

U

Uenr

U

Uenr

Uenr

MM

r

M

r

M

r

M

r

8813.0, Uenr

Page | 156

Therefore,

195.0*881.0002.0*882.0

0071.0*882.0002.0*882.06075.0 2

enrfR

Design 2

In Design 2, the depleted UO2 fuel region is inscribed inside a thin enriched fuel annulus. The mass of U-235 in the depleted uranium pin is:

mdep,U-235 = Udep

OdepU

depU

Udepfdepf rMM

MLR ,

16

,

2

*2

The mass of U-235 in the enriched fuel annulus is:

menr,U-235 = Uenr

OenrU

enrUUenrfdepff r

MM

MLRR ,

16

,

22

*2)(

Again, to calculate the thickness of the thin enriched fuel layer in Design 2, a simple mass balance relation can be formulated based on the mass of U-235 in the reference CANDU-6 and Design 2 fuel pins.

mnat,U-235 = mdep, U-235 + menr, U-235

Uenr

OenrU

enrUUenrfdepff

Udep

OdepU

depU

UdepfdepfUnat

OnatU

natUUnatff

rMM

MLRR

rMM

MLRr

MM

MLR

,

16

,

22

,

16

,

2

,

16

,

2

*2)(

*2*2

Based on the previous terminology for Aenr,U, Adep,U and Anat,U, the above equation can be simplified

to:

UenrUenrUenrfdepffUdepUdepUdepfdepfUnatUnatUnatff rLRRrLRrLR ,,,

22

,,,

2

,,,

2 )(

UenrUenrUenrdepffUdepUdepUdepdepfUnatUnatUnatf rRRrRrR ,,,

22

,,,

2

,,,

2 )(

UnatUnatUnatfUenrUenrUenrfUdepUdepUdepdepfUenrUenrUenrdepf rRrRrRrR ,,,

2

,,,

2

,,,

2

,,,

2

Renrf = 0.098 cm ≅ 0.1 cm

Page | 157

Separating Rdepf to one side gives:

UdepUdepUdepUenrUenrUenr

UnatUnatUnatUenrUenrUenrf

depfrr

rrRR

,,,,,,

,,,,,,

2

2

UdepUdepUdepUenrUenrUenr

UnatUnatUnatUenrUenrUenrf

depfrr

rrRR

,,,,,,

,,,,,,

2

(iii)

Again, assuming the densities do not change for different enrichments:

UnatUdepUenr ,,,

Eq. (iii) simplifies to:

UdepUdepUenrUenr

UnatUnatUenrUenrf

depfrr

rrRR

,,,,

,,,,

2

Evaluating the above equation gives:

002.0*8815.0195.0*8813.0

0071.0*8815.0195.0*8813.06075.0 22

cmRdepf

cmcmRdepf 600.05994.0

Thickness of the enriched layer then becomes:

Design 3

Assumptions:

Density of uranium metal, ρdep,U = ρenr,U = 18.95 g/cm3

In Design 3, the depleted U-metal fuel region is inscribed inside a thin enriched U-metal fuel annulus. Furthermore, uranium metal is used in the modified fuel bundle for both depleted and enriched regions. The mass of U-235 in the depleted uranium pin is:

tenrf =0.6075 cm – 0.6 cm = 0.0075 cm

Page | 158

mdep,U-235 = Udep

depU

depU

Udepfdepf rM

MLR ,,

2

The mass U-238 in the depleted uranium pin is:

mdep,U-238 = )1( ,,

2

Udep

depU

depU

Udepfdepf rM

MLR


menr,U-235 = Uenr

enrU

enrUUenrfdepfm r

M

MLRR ,,

22 )(


menr,U-238 = )1()( ,,

22

Uenr

enrU

enrUUenrfdepfm r

M

MLRR

where Rm is the new fuel pellet radius. Again, to calculate the thickness of the thin enriched U-metal fuel layer in Design 3, a simple mass balance relation can be formulated based on the mass of U-235 in the reference CANDU-6 and Design 3 fuel pins.


Uenr

enrU

enrUUenrfdepfm

Udep

depU

depU

UdepfdepfUnat

OnatU

natUUnatff

rM

MLRR

rM

MLRr

MM

MLR

,,

22

,,

2

,

16

,

2

)(

*2

Based on the previous terminology for Anat,U, the above equation can be simplified to:

UenrUenrfdepfmUdepUdepfdepfUnatUnatUnatff rLRRrLRrLR ,,

22

,,

2

,,,

2 )(

The above equation can be further simplified by multiplying both sides by πLf.

UenrUenrdepfmUdepUdepdepfUnatUnatUnatf rRRrRrR ,,

22

,,

2

,,,

2 )(

UnatUnatUnatfUenrUenrmUenrUenrUdepUdepdepf rRrRrrR ,,,

2

,,

2

,,,,

2 )( (iv)

Similarly, a simple mass balance relation can be formulated based on the mass of U-238 in the reference CANDU-6 and Design 3 fuel pins.


Page | 159

)1(*2

,

16

,

2

Unat

OnatU

natUUnatff r

MM

MLR

= )1( ,,

2

Udep

depU

depU

Udepfdepf rM

MLR

+

)1()( ,,

22

Uenr

enrU

enrUUenrfdepfm r

M

MLRR

)1()()1()1( ,,

22

,,

2

,,,

2

UenrUenrfdepfmUdepUdepfdepfUnatUnatUnatff rLRRrLRrLR

The above equation can be further simplified by multiplying both sides by πLf.

)1()()1()1( ,,

22

,,

2

,,,

2

UenrUenrdepfmUdepUdepdepfUnatUnatUnatf rRRrRrR

)1(

)1()]1()1([

,,,

2

,,

2

,,,,

2

UnatUnatUnatf

UenrUenrmUenrUenrUdepUdepdepf

rR

rRrrR

(v)

Eq. (iv) and Eq. (v) can be solved for the two unknown Rdepf and Rm by multiplying Eq. (iv) by

Uenr

Uenr

r

r

,

, )1( and subtracting it from Eq. (v). This gives:

Uenr

Uenr

UnatUnatUnatf

UenrUenrm

Uenr

Uenr

UenrUenrUdepUdepdepf

r

rrR

rRr

rrrR

,

,

,,,

2

,,

2

,

,

,,,,

2

)1(

)1()1(

)(

(vi)

Subtracting Eq. (v) from Eq. (vii) gives:

)1()1(

)]1()1([)1(

)(

,,,

2

,

,

,,,

2

,,,,

2

,

,

,,,,

2

UnatUnatUnatf

Uenr

Uenr

UnatUnatUnatf


Uenr

Uenr


rRr

rrR

rrRr

rrrR

))1()1(

(

)1()1()1()1(

,

,

,

,,,

2

,,,,,,

,

,

,,

2

Unat

Uenr

Uenr

UnatUnatUnatf

UenrUenrUdepUdepUenrUenr

Uenr

Uenr

UdepUdepdepf

rr

rrR

rrrr

rrR

Page | 160

)1()1(

)1()1(

,

,

,

,,,

2

,

,

,

,,

2

Unat

Uenr

Uenr

UnatUnatUnatf

Udep

Uenr

Uenr

UdepUdepdepf

rr

rrR

rr

rrR

)1()1(

)1()1(

,

,

,

,,

,

,

,

,,,

2

2

Udep

Uenr

Uenr

UdepUdep

Unat

Uenr

Uenr

UnatUnatUnatf

depf

rr

rr

rr

rrR

R

)1()1(

)1()1(

,

,

,

,,

,

,

,

,,,

2

Udep

Uenr

Uenr

UdepUdep

Unat

Uenr

Uenr

UnatUnatUnatf

depf

rr

rr

rr

rrR

R


1772.07556.18

3229.3

)002.01(195.0

)195.01(002.095.18

)0071.01(195.0

)195.01(0071.08815.0*6.106075.0 2

depfR

cmRdepf 4209.0

Thickness of the depleted Uranium metal layer would be 0.4209 cm. Substituting this value back into Eq. (iv) gives:

UnatUnatUnatfUenrUenrmUenrUenrUdepUdepdepf rRrRrrR ,,,

2

,,

2

,,,,

2 )(

)( ,,,,

2

,,,

2

,,

2

UenrUenrUdepUdepdepfUnatUnatUnatfUenrUenrm rrRrRrR

UenrUenr

UenrUenrUdepUdepdepfUnatUnatUnatf

mr

rrRrRR

,,

,,,,

2

,,,

2

2)(

Page | 161

UenrUenr

UenrUenrUdepUdepdepfUnatUnatUnatf

mr

rrRrRR

,,

,,,,

2

,,,

2 )(


195.0*95.18

)195.0*95.18002.0*95.18(*4209.00071.0*8815.0*6.10*6075.0 22 mR

18196.06953.3

)6479.0(02448.0

mR

mR 0.4266 cm

Thickness of the enriched layer then becomes:

tenrf =0.4266 cm – 0.4209 cm = 0.00567 cm