Nuclear Science
2013

Minor Actinide Burningin Thermal Reactors

A Report by the Working Party on
Scientific Issues of Reactor Systems

1.00E+10

UO2 spent fuel1.00E+09

Pu single recycle
TRU single recycle

Radiotoxicity (Sv/tHMi)

1.00E+08

1.00E+07

1.00E+06

1.00E+05

1.00E+04

1.00E+03
1.E+00

1.E+01

1.E+02

1.E+03

1.E+04

1.E+05

1.E+06

1.E+07

1.E+08

Decay time (years)

NEA

Nuclear Science

Minor Actinide Burning in Thermal Reactors
A Report by the Working Party on
Scientific Issues of Reactor Systems© OECD 2013
NEA No. 6997
NUCLEAR ENERGY AGENCY
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Cover photo: The Minor Actinide Laboratory at the EC Institute for Transuranium Elements in Germany (European Communities).

FOREWORD

Foreword

The actinides (or actinoids) are those elements in the periodic table from actinium
upwards. Uranium (U) and plutonium (Pu) are two of the principal elements in nuclear
fuel that could be classed as major actinides. The minor actinides are normally taken to
be the triad of neptunium (Np), americium (Am) and curium (Cm). The combined massesof the remaining actinides (i.e. actinium, thorium, protactinium, berkelium, californium,
einsteinium and fermium) are small enough to be regarded as very minor tracecontaminants in nuclear fuel. Those elements above uranium in the periodic table are
known collectively as the transuranics (TRUs).
The operation of a nuclear reactor produces large quantities of irradiated fuel(sometimes referred to as spent fuel), which is either stored prior to eventual deep
geological disposal or reprocessed to enable actinide recycling. A modern light water
reactor (LWR) of 1 GWe capacity will typically discharge about 20-25 tonnes of irradiatedfuel per year of operation. About 93-94% of the mass of uranium oxide irradiated fuel iscomprised of uranium (mostly 238U), with about 4-5% fission products and ~1% plutonium.
About 0.1-0.2% of the mass is comprised of neptunium, americium and curium. Theselatter elements accumulate in nuclear fuel because of neutron captures, and they
contribute significantly to decay heat loading and neutron output, as well as to theoverall radiotoxic hazard of spent fuel. Although the total minor actinide mass isrelatively small – approximately 20-25 kg per year from a 1 GWe LWR – it has adisproportionate impact on spent fuel disposal, and thus the longstanding interest intransmuting these actinides either by fission (to fission products) or neutron capture inorder to reduce their impact on the back end of the fuel cycle. The combined masses of
the trace actinides actinium, thorium, protactinium, berkelium and californium in
irradiated LWR fuel are only about 2 parts per billion, which is far too low for them to beconsidered candidates for transmutation.
Fast reactors are needed to transmute TRUs because fast neutron cross sections are
generally more effective in the fissioning of TRUs. However, studies have demonstrated
that TRU transmutation rates can also be achieved in thermal reactors, although withserious limitations due to their accumulation through recycling and their impact on thesafety of the plants.
The transmutation of TRUs could potentially be carried out in many thermal reactors
operating today, while waiting for a similar programme in fast reactors to allow
commercial-scale operations in 20 to 30 years or more. Investment in fuel cycle plants
could lead to even more efficient transmutation in fast reactors towards the end of the
century. In the interim, the potential contribution of thermal reactors should not be
overlooked.
A considerable amount of research has already been conducted on minor actinide
transmutation in thermal reactors, and the purpose of this report is to summarise thefindings of this research. The report concentrates on general conclusions related tothermal reactors and foregoes a lengthy examination of the more technical details. Whilea commercial-scale implementation programme for a specific reactor type will need toaddress a multitude of very specific questions, the objective of this report is to provide
the broad understanding necessary to inform high-level strategy and decision making.

MINOR ACTINIDE BURNING IN THERMAL REACTORS, NEA No. 6997, © OECD 2013

3

FOREWORD

Chapter 1 of the report provides an introduction to minor actinide nuclear propertiesand discusses some of the arguments in favour of minor actinide recycling. The
introduction is not specific to thermal reactors but could apply to any nuclear system ingeneral. Chapter 2 discusses the potential role of thermal reactors in minor actinide
recycling; Chapter 3 looks at the various technical issues and challenges presented byminor actinide recycling; Chapter 4 examines fuel cycle issues; Chapter 5 presents
implications for thermal reactor operations, fuel design, core management, andsafety/dynamics responses; Chapter 6 explores operations and safety issues; Chapter 7
considers the economics of minor actinide recycling and attempts to make somequantitative estimates insofar as current knowledge permits; Chapter 8 discusses
research and development needs and Chapter 9 summarises findings, making
recommendations for the direction of future R&D efforts.

Acknowledgements

The NEA Secretariat is grateful to the members of the Working Party on ScientificIssues of Reactor Systems (WPRS) who participated in meetings and helped to improve
this report. Contributions from the following experts in particular are greatly appreciated:
K. Hesketh (NNL), D. Porsch (AREVA), G. Rimpault (CEA), T. Taiwo (ANL), A. Worrall (ORNL).

4MINOR ACTINIDE BURNING IN THERMAL REACTORS, NEA No. 6997, © OECD 2013

TABLE OF CONTENTS

Table of contents

1. Introduction ......................................................................................................................................................... 9
1.1 Neptunium, americium and curium nuclear properties .................................................................... 9
1.2 Rationale for minor actinide transmutation ........................................................................................ 13
1.2.1 Radiotoxicity reduction ................................................................................................................... 13
1.2.2 Decay heat reduction ....................................................................................................................... 15
1.3 Minor actinide transmutation strategies .............................................................................................. 15
1.3.1 Homogeneous/heterogeneous recycling ..................................................................................... 15
1.3.2 Single/multiple recycling ................................................................................................................ 16
1.3.3 Storage ................................................................................................................................................ 17
1.3.4 Minor actinide recycling in thermal reactors versus fast reactors ........................................17
1.3.5 Minor actinide recycling in thermal reactors as a technology demonstrator ....................18
2. Potential role and objectives of minor actinide recycling in thermal reactors................................21
2.1 Light water reactor (LWR) studies .......................................................................................................... 21
2.2 Heavy water reactor (HWR) studies ....................................................................................................... 22
2.3 High-temperature reactor (HTR) studies .............................................................................................. 23
2.4 Survey of experimental studies .............................................................................................................. 24
3. Issues associated with utilisation of MA-TRU fuels...............................................................................29
3.1 Recycling modes – homogeneous and heterogeneous ......................................................................29
3.2 Reprocessing ............................................................................................................................................... 30
3.3 Separation technologies ........................................................................................................................... 31
3.4 Fuel manufacturing ................................................................................................................................... 33
3.5 Fresh fuel transport and handling ......................................................................................................... 34
3.6 Irradiation testing ...................................................................................................................................... 34
3.7 Licensing ...................................................................................................................................................... 35
3.8 Irradiated fuel ............................................................................................................................................. 35
3.9 Waste ............................................................................................................................................................ 36
3.10 Impact on geological disposal ............................................................................................................... 36
3.11 Overall timescales .................................................................................................................................... 36
3.12 Non-technical considerations ............................................................................................................... 36
4. Fuel cycle issues ............................................................................................................................................... 39
4.1 Source terms ............................................................................................................................................... 39
4.2 Fuel fabrication ........................................................................................................................................... 42
4.3 Transport ..................................................................................................................................................... 42
4.4 Utilisation rates .......................................................................................................................................... 43
4.5 Irradiated fuel inventories ....................................................................................................................... 44
4.6 Radiotoxicity and environmental impact ............................................................................................. 44

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TABLE OF CONTENTS

5. Fuel and core design ....................................................................................................................................... 45
5.1 Fuel design ................................................................................................................................................... 45
5.2 Core design .................................................................................................................................................. 45
5.2.1 Equilibrium UO2 core........................................................................................................................ 46
5.2.2 Equilibrium Am-Cm core ................................................................................................................ 46
5.2.3 Nuclear design parameters ............................................................................................................ 47
5.3 Material balance ......................................................................................................................................... 51
5.4 Fuel matrices ............................................................................................................................................... 52
6. Operations, safety and licensing ................................................................................................................. 55
6.1 Initial fissile loading .................................................................................................................................. 55
6.2 Impact on plant operations ..................................................................................................................... 56
6.2.1 Fresh fuel receipt .............................................................................................................................. 56
6.2.2 Core loading ....................................................................................................................................... 56
6.2.3 Core operation ................................................................................................................................... 57
6.3 Code validation ........................................................................................................................................... 57
6.4 Licensing timescales.................................................................................................................................. 57
7. Economics .......................................................................................................................................................... 59
7.1 Costs and benefits ...................................................................................................................................... 59
7.2 Cost impact.................................................................................................................................................. 60
7.2.1 Separation............................................................................................................................ 60
7.2.2 Transport ............................................................................................................................................ 61
7.2.3 Fresh fuel receipt and storage........................................................................................................ 61
7.2.4 Core reactivity effect ........................................................................................................................ 61
7.2.5 Licensing ............................................................................................................................................. 62
7.2.6 Discharge, storage and transport of spent MA-TRU fuels .......................................................62
7.2.7 Spent fuel management.................................................................................................................. 62
7.2.8 Overall cost impact........................................................................................................................... 62
7.3 Economic benefits ...................................................................................................................................... 62
7.3.1 Repository cost savings ................................................................................................................... 62
7.3.2 Avoidance of radiological doses .................................................................................................... 63
7.4 Subsidisation mechanisms ...................................................................................................................... 63
8. Research and development needs .............................................................................................................. 65
8.1 Separations .................................................................................................................................................. 65
8.2 Fuel fabrication ........................................................................................................................................... 65
8.3 Fuel transport.............................................................................................................................................. 65
8.4 Fuel design ................................................................................................................................................... 66
8.5 Core design .................................................................................................................................................. 66
8.6 Spent fuel characterisation ...................................................................................................................... 66
8.7 Fuel cycle assessment ............................................................................................................................... 66
8.8 Overall timescales ...................................................................................................................................... 67
9. Summary and recommendations ............................................................................................................... 69
References ............................................................................................................................................................... 71
List of bibliographic references related to the Report on Minor Actinide Burning in Thermal
Reactors ................................................................................................................................................................... 73

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TABLE OF CONTENTS

List of figures
1.1: Radiotoxicity profiles for LWR UO2, LWR MOX and LWR TRU recycling............................. 15
4.1: CORAIL PWR fuel assembly....................................................................................................... 39
4.2: Decay heat, gamma and neutron sources during fuel fabrication for equilibrium
isotopic compositions ................................................................................................................ 41
4.3: Radiotoxicity versus decay time for CORAIL fuel variants (based on [9]) ........................... 44
5.1: Material balance for americium and curium target rods ...................................................... 52

List of tables
1.1: Decay properties of neptunium americium and curium .................................................... 11
1.2: Neutron interaction cross-sections for Np, Am and Cm weighted with LWR neutron
spectrum (barns) ...................................................................................................................... 13
2.1: Summary of historical studies of minor actinide recycling in thermal reactors............. 24
2.2: Summary of minor actinide fuel experiments ..................................................................... 27
4.1: Equilibrium isotopic compositions for multi-recycling scenarios from [9] ...................... 40
4.2: Decay heat, gamma and neutron sources during fuel fabrication for different TRU
isotopic compositions .............................................................................................................. 41
4.3: Dose rates (mSv/h) ................................................................................................................... 42
5.1: Isotopic composition of Am-Cm feed mix ............................................................................ 47
5.2: Impact of Am-Cm target loading on equilibrium nuclear design parameters ................ 48
5.3: HFP radial peaking factor F∆H .................................................................................................. 49
5.4: HFP moderator temperature coefficient (pcm/°C) ............................................................... 50
5.5: HZP moderator temperature coefficient (pcm/°C) ............................................................... 50
5.6: Doppler defects from HFP to HZP in pcm ............................................................................. 50
5.7: HFP boron reactivity coefficients and nuclear design limits (pcm/ppm) ......................... 50
5.8: HZP boron reactivity coefficients and nuclear design limits (pcm/ppm) ......................... 50
5.9: HZP control rod worths in pcm (minus the highest worth rod)......................................... 51
5.10: Best-estimate shut-down margins (pcm) ............................................................................. 51
5.11: HZP delayed neutron fraction ................................................................................................ 51
5.12: Boration limits .......................................................................................................................... 51
6.1: Licensing of MA/TRU fuel ........................................................................................................ 58

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1. INTRODUCTION

1. Introduction

1.1 Neptunium, americium and curium nuclear properties
It would be helpful to start by reviewing the basic nuclear decay and cross-section
properties of Np, Am and Cm, because this will highlight the reasons why these elements
are considered a challenge in the nuclear fuel cycle. Table 1.1 lists the nuclear decay
properties of the main neptunium, americium and curium isotopes, including their halflives, decay modes, branching ratios, gamma emissions, effective dose coefficients, total
neutron emission rates and decay heat outputs.
Of the TRUs in irradiated nuclear fuel, neptunium, americium and curium are thethree that are usually considered as candidates for partitioning and transmutation (P&T).
Although there are other TRUs present, these three are dominant in terms of mass.
Irradiated LWR fuel typically contains, about 600 g/tHM of neptunium; 850 g/tHM ofamericium and about 50 g/tHM of curium at 10 years after discharge. Because the 241Am
arises from 241Pu decay and curium isotopes are relatively short lived, these proportions
vary depending on the cooling time after discharge and the discharge burn-up.
Neptunium, which is predominantly represented by the single isotope 237Np, is asignificant contributor to long-term radiotoxicity, because of its very long half-life.
However, 237Np does not contribute significantly to decay heat output.
Americium is generally considered as a prime candidate for transmutation because itis present in relatively large amounts and is a significant contributor to gamma activity
and radiotoxicity, especially after about 500 years cooling time when the contribution offission products has decreased by several orders of magnitude. Americium hasreasonably large nuclear cross-sections and is amenable to destruction in an intense
neutron flux by a combination of neutron captures and fissions. In irradiated nuclear fuel,241
Am is the dominant nuclide, though there are small but significant quantities of 242Am,
242M
Am and 243Am.
Curium makes a significant contribution to gamma activity and radiotoxicity and isalso a major contributor to neutron emissions. Curium is not well suited totransmutation, because the fission and capture cross-sections of the principal isotopes(242Cm and 244Cm) are quite low and it is not usually possible to transmute themeffectively. Although 242Cm has a very short half-life (163 days), it is continually generated
in irradiated fuel from the decay of 242Am (141 year half-life).
The relatively short half-life of 244Cm (18.1 years) suggests that interim storage may
perhaps be a better strategy for this isotope, allowing time for it to decay naturally before
being incorporated in recycling targets. Such a decay storage strategy would help reduce
the strong heat source from 244Cm in fabricating the target fuels. Reducing the mass of244
Cm in the target fuel would also reduce the production rate of 245Cm. Although much of
the 245Cm produced during irradiation of target fuels in a thermal reactor undergoes
fission, that which does remain will be problematic because of its 8 500-year half-life.
One rationale for irradiating Cm and not separating it, is that for some of the proposed
separation processes it is not possible to separate Am and Cm individually and therefore
if Am is to be recycled, then Cm must be as well.

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1. INTRODUCTION

The decay modes are included principally because many of these nuclides undergo
spontaneous fissions that are accompanied by neutron emissions, which determine theshielding requirements for storage, transport and handling. The spontaneous fission rate
for 237Np is zero. The spontaneous fission rates of the americium isotopes are all very low
and although the typical Am-Cm mix from irradiated LWR fuel contains about 90% Am,
the 10% or so curium content dominates neutron emissions. This is mainly due to thecontribution of 244Cm, which is the predominant curium isotope and has a short half-lifeand a relatively high spontaneous fission branching ratio. In addition to spontaneous
fission neutrons, there is also a neutron source from (α,n) reactions.
Gamma emissions are difficult to characterise, because in some cases there are
multiple lines (e.g. there are 129 lines for 241Am). Table 1.1 identifies the gamma
emissions with the highest intensity (which usually tends to be at low energy) and also
the line (or lines) with the highest intensities at high energies. These high-energy gamma
emissions are the most difficult to shield and are important in defining the shielding
requirements.
Direct gamma emissions from 237Np are relatively mild because of the low decay rate
and the low energy of the highest intensity emissions. However, its short-lived daughter
233
Pa is in secular equilibrium with 237Np and has an energetic gamma of 312 keV atintensity 0.386 that dominates overall gamma emissions from the 237Np decay chain.
241
Am has a 60 keV line at 0.359 intensity. 243Am is a strong gamma emitter, witha75 keV line at 0.674 intensity. Moreover, 239Np, which is a short-lived daughter of 243Am
decays, has a very energetic gamma line at 278 keV of 0.144 intensity. The shorter halflife of 241Am and its higher abundance means that 241Am is the dominant contributor.
Gamma emissions from curium are dominated by 244Cm.

The last column in Table 1.1 indicates total neutron emissions from the oxide, which
combines spontaneous neutron emissions with α-n emissions. For 241Am and 243Am, α-nemissions dominate over spontaneous fissions (the figures quoted in Table 1.1 give thetotal neutron emissions for the oxide), but for 242Cm and 244Cm the spontaneous neutron
source dominates. Since Cm is the dominant neutron emitter in the Am-Cm mix, theform of the material (whether metal, oxide, carbide or nitride) does not significantly
affect neutron emissions.
An important consideration for neptunium, americium and curium is radiotoxicity.
The ingestion effective dose coefficients as listed in ICRP-72 [1] are the same for all theisotopes listed in Table 1.1 (5.0E-4 Sv/Bq) so that their direct contributions to radiotoxic
potential are proportional to the decay rates. Because of its very long half-life, 237Np only
contributes significantly to total radiotoxicity at very long cooling times. 241Am is thedominant contributor of the americium isotopes, due to its higher abundance and its
half-life being shorter than 243Am. 244Cm is the highest contributor of the Cm isotopes, due
to its abundance and short half-life. The total radiotoxicity depends on the ingestion
effective dose coefficients of the respective decay chains (i.e. contributions both from thedecay of the parent nuclide and all the daughters in its decay chain), but the various
decay chains are not overly dissimilar in this respect.
The contribution of americium and curium to radiotoxicity decays by a factor of about
2 after 100 years, 10 after 1 000 years, and then falls to about 0.5% of its initial value after
10 000 years. At cooling times of over 1 000 years, the long-lived nuclides, 243Am and 245Cm
are the dominant contributors. This decay profile is much slower than the decay profile
of fission products and with full recycling of uranium and plutonium would become thelimiting factor controlling the radiotoxic content of a repository at 1 000 years.

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1. INTRODUCTION

Table 1.1: Decay properties of neptunium americium and curium

NuclideHalf-life

Decay modes

Branching ratios

237Np

2.14E6
years

Alpha

1.0 α

239Np

2.35 days

241Am

432 years

Principal gamma
emissions

29 keV intensity
0.153; 86 keV
intensity 0.123

Neutronemissions(spontaneous +alpha-n fromoxide) n/sec/gHM

Decay
heat
(W/kgHM)

351.4

2.07E-2

2752

114.7

278 keV intensity
0.144
Alpha and spontaneous
fission (+neutron
emissions)

1.0 α
3.77E-12 SF

60 keV intensity
0.359; 662 keV
intensity 3.6E-6

242Am

16.02 hours

242MAm

141 years

Internal decay; alpha andspontaneous fission(+neutron emissions)0.9955 internal
decay; 0.0045 α;
1.6E-10 SF

49 keV intensity
1.9E-3; 163 keV
intensity 2.3E-4

186.6

4.5

243Am

7370 years

Alpha and spontaneous
fission (+neutron
emissions)

1.0 α

75 keV intensity
0.674; 662 keV
1.1E-5138.6

6.4

44 keV at 3.25E-4
intensity; two at
up to 942 keV
combined
intensity 6E-7

2.360E7121228.0

278 keV at 0.14
intensity; five up
to 755 keV
combined
intensity 8.5E-5

4.869E4

1860.743 keV at 2.5E-4
intensity; 818 keV
at 7.2E-7 intensity

1.092E7

2841.8

242Cm

243Cm

244Cm

163 days29.1 years

18.1 years

Not relevant because of short half-life

Alpha and spontaneous
fission (+neutron
emissions)

Alpha and positronemission (β+)

Alpha and spontaneous
fission (+neutron
emissions)

3.7E-11 SF
1.0 α
6.33E-8 SF

0.9976 α
0.0024 β+

1.0 α
1.35E-6 SF

245Cm

8 500 years

Alpha

1.0 α

175 keV at 0.095
intensity

125.3

5.8

246Cm

4 730 years

Alpha and spontaneous
fission (+neutron
emissions)

1.0 α

45 keV at 2.8E-4
intensity

8.759E6

10.2

2.61E-4 SF

Source: JEF-PC nuclear data program, OECD-NEA; neutron emissions from FISPIN inventory
programme.

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1. INTRODUCTION

Because of the gamma and neutron emissions, handling Am and Cm demands
substantially more shielding than is required for MOX fuel fabrication. For example, at
the Institute for Transuranium Elements (ITU) in Karlsruhe, Germany, the Minor Actinide
Laboratory used to fabricate small quantities of Am and Cm into targets has 50 cm ofwater shielding and 5 cm of lead shielding at working level. This is less than the shielding
requirement for irradiated fuel, but substantial nonetheless. At the ITU Minor Actinide
Laboratory, normal operations involving Am and Cm samples are carried out using
remote manipulators, supported with remote control and robotics. The cells are, however,standard gloveboxes that allow manual intervention in the absence of radioactive
sources, as required for maintenance and repair.
Table 1.2 shows a sample of neutron cross-sections for the various Np, Am and Cm
isotopes. It shows 3-group fission and capture cross-sections weighted with an LWR
neutron spectrum, in the fast, epithermal and thermal energy ranges. These crosssections are only intended for illustration and in practice will vary depending on theinitial enrichment and on the burn-up, as well as the specific LWR type. The data do notinclude minor cross-sections such as multiple neutron events (n, 2n) or (n, 3n).
The epithermal cross-sections include resonance self-shielding, which is highly
dependent on the specific reactor conditions. In particular, the resonance region crosssections depend on the physical form of the minor actinide or transuranic (MA-TRU) fuel,whether inhomogeneous or homogeneous, for which different resonance self-shielding
factors would be expected to apply.
The ratios of fission to capture are indicative of whether the nuclides will primarily
undergo fission or neutron captures as shown in Table 1.2. This table shows the fast,epithermal and thermal cross-sections for fission and capture when weighted by LWR
neutron energy spectrum. 241Am is more complicated because neutron captures may leadto either the ground state of 242Am or its metastable state. In order to clearly demonstrate
this complexity, the data for 241Am have been split over three rows, one for
241
Am(n,γ)242Am, one for 241Am(n,γ)242MAm and one for 241Am fission.
In the case of 237Np, the dominant nuclear reaction over the three neutron groups is
capture leading to 238Np. However, 238Np is predominantly removed by fissions. The endpoint is the destruction of 237Np by fissions via two neutron interactions, with the firstneutron capture as the rate-controlling step.
241
Am also predominantly undergoes neutron capture, transmuting either to 242Am or
Am (the branching ratio favours the ground state 242Am). The 242Am can then be
removed by fission or capture, with fission largely dominating for the 242MAm branch, butslightly less so for the ground state branch. In either case, removal of 241Am is via twoneutron reactions, the first again being rate limiting.
242M

Thermal captures also dominate over fissions for 242Cm and 244Cm, though for 243Cm
and 245Cm the thermal fissions dominate. The overall trend is that neutron captures tend
to predominate in a thermal neutron spectrum, building up higher nuclides in theprocess. Fast neutron interactions make only a minor contribution to the overall reaction
rate, because the cross-sections are much lower.
In terms of the core design behaviour, the cross-sections in Table 1.2 suggest that
although neptunium, americium and curium contribute limited useful energy output
from fissions in a thermal reactor, captures are nevertheless significant. Therefore, their
behaviour in a thermal reactor is to some extent analogous to that of burnable poisons.

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1. INTRODUCTION

Table 1.2: Neutron interaction cross-sections for Np, Am and Cm weighted with
LWR neutron spectrum (barns)
NuclideFastfissionFast capture

Epithermal
fissionEpithermal
captureThermalfission

ThermalcaptureDominant
reaction

237Np

0.86

0.62

0.02

50.49

0.01

164.50Capture

238Np

2.34

0.07

74.43

8.46

934.9

93.74

Fission

241Am to 242Am

-

0.73

-

75.37

-

654.70241Am to242MAm

-

0.14

-

11.26

-

72.75

241Am fission

0.86

-

0.60

-

4.98

-

242Am

2.19

0.33

99.49

29.97

1087.00

2423.00

Fission

242MAm

2.24

0.23

118.0017.44

4713.00

1106.00

Fission

243Am

0.71

0.72

0.09

143.31

0.03

43.71

Capture

242Cm

1.18

0.17

0.33

9.48

2.81

7.60

Capture

243Cm

2.48

0.07

150.0024.63

233.4048.44

Fission

244Cm

1.07

0.28

0.67

36.89

0.49

7.16

Capture

245Cm

2.12

0.13

54.93

10.00

784.90119.80Fission

246Cm

0.00

0.10

0.00

7.31

0.00

0.67

CaptureCapture1.2 Rationale for minor actinide transmutation
1.2.1 Radiotoxicity reduction
The black points in Figure 1.1 plot the total ingestion radiotoxicity (in Sieverts) of
1 tonne of spent LWR UO2 fuel, irradiated to a burn-up 45 GWd/t as a function of coolingtime after discharge. The radiotoxicity decay profile can be divided crudely into threesegments:
The first segment, which extends up to a cooling time of ~500 years, is dominated by
the radiotoxicity of the fission products. These are mostly relatively short-lived, except
for a small contribution from some very long–lived fission products such as 129I. The
short-lived fission products dominate the total radiotoxicity up until about 500 years,
after which time all the main contributors have decayed. The second segment, which
covers decay times between 1 000 years and 105 years is dominated by the contribution of
plutonium and the minor actinides, which by the end of this period have essentially all
decayed. Finally, in the segment after 105 years, the radiotoxicity is determined by thevery long-lived fission products and 237Np.
The red points in Figure 1.1 plot the equivalent radiotoxicity decay profile, but in thiscase assuming a reprocessing cycle in which the plutonium has been recycled a single
time as MOX fuel. The MOX recycling case data have been generated in such a way thatthe radiotoxicities per unit of useful electrical output are comparable so as to provide afair comparison. The radiotoxicity of this single recycling MOX scenario is initiallyslightly higher than the equivalent UO2 fuel, because the MOX fuel inventory has a higher
content of minor actinides. For longer periods up to ~105 years, however, the MOX case is

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1. INTRODUCTION

slightly lower because some of the plutonium has been destroyed by fissions. Overall, theradiotoxicities of the UO2 and single recycling MOX scenarios are comparable.
It may be technically possible to design geological repositories that will isolate thebulk of the radionuclides from the biosphere for hundreds or even thousands of years.
The design objective would be initially to contain the radionuclides for perhaps several
hundred years and then rely on engineered and natural barriers to delay the subsequenttransport of mobile radionuclides out of the repository and into the biosphere until
perhaps several thousand years later. But no conceivable repository design could beguaranteed to isolate the radionuclides over the geological time scale of 1 million years or
more, which would be needed for complete decay. Therefore, repository designers startwith the premise that on geological timescales there will be non-containment of theradionuclides, but that the design should aim to delay the breakdown of containment aslong as possible.
When discussing the design of a repository, there are two distinct (but linked) aspects.
One is the purely technical assessment and the second is the presentation to the public.
The technical assessment will necessarily involve complex analysis such as the migration
rate of radionuclides and peak dose uptake over very long timescales (up to 1 million years),
which are not readily understandable to the general public. For periods up to about
1 000 years, the general public might reasonably be expected to understand that engineered
solutions might be effective, because historic structures have survived for such periods andlonger. But for longer periods, abstract technical arguments are likely to be much lesspersuasive.
It would be more persuasive if the total radiotoxicity could be brought down to thelevel of the third segment on the more understandable 1 000-year timescale. The blue
points in Figure 1.1 plot how this might be achieved, by separating out the minor
actinides and transmuting them. This has been done in a somewhat artificial way here,by separating all the TRU from the 1 tonne of spent UO2 fuel and assuming they areirradiated to almost complete destruction in a thermal reactor spectrum. Although this is
not a rigorous approach, it serves here to indicate what might be achievable in an ideal
scenario and therefore indicates the maximum extent to which minor actinide burning
could reduce the radiotoxicity. In this case, the total radiotoxicity reaches a much lower
plateau in the second segment, approaching that of the third (long-term) segment. This is
an important part of the rationale for minor actinide partitioning (separation) andtransmutation, because it brings forward the long-term radiotoxicity plateau towards the
1 000 year timescale, which the general public could better relate to.
This same argument has often been expressed in a different way by regarding theradiotoxicity of the original uranium ore prior to extraction as a target threshold for therepository to meet. The argument is that once the radiotoxicity of the spent fuel has
decayed to below that of the uranium ore, then this is a convenient level at which future
concern can be regarded as no longer significant. While this argument has some merit,especially when applied for comparative purposes, it is nevertheless somewhat arbitrary
because although the radiotoxicity matches that of the original ore, it is difficult to argue
that the geological environment of the repository exactly matches that of the original ore
body. For this reason, this report does not to use this threshold definition, but rather
regards minor actinide transmutation as a possible means of accelerating the timescale
during which the radiotoxicity decreases.
Although radiotoxicity is a very important parameter, its impact on the performanceof a geological repository is limited. A recent OECD/NEA study [2] has highlighted that theremoval of minor actinides from the inventory of a geological repository has very little
impact on peak radiological environmental dose. This is because, in the repository
environments, the minor actinides tend to have low mobility and environmental dose iscontrolled instead by volatile elements. However, the study does note that removal ofminor actinides does have a major beneficial effect in intrusion scenarios. Intrusion

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1. INTRODUCTION

scenarios involve penetration of waste packages by deep drilling activities, where there is
the potential of the drilling operators receiving significant radiological doses from theminor actinides, depending on the specific scenario postulated.
All scenarios for minor actinide transmutation are dependent on the development ofeffective chemical separation methods for separating minor actinides from fission
products, uranium and plutonium, with high separation factors being achievable. The
presence of significant amounts of minor actinides in fission product streams has thepotential to undermine the justification for minor actinide transmutation strategies.

Figure 1.1: Radiotoxicity profiles for LWR UO2, LWR MOX and LWR TRU recycling
1.00E+10

UO2 spent fuel1.00E+09

Pu single recycle
TRU single recycle

Radiotoxicity (Sv/tHMi)

1.00E+08

1.00E+07

1.00E+06

1.00E+05

1.00E+04

1.00E+03
1.E+00

1.E+01

1.E+02

1.E+03

1.E+04

1.E+05

1.E+06

1.E+07

1.E+08

Decay time (years)

1.2.2 Decay heat reduction
The principal contributors to decay heat from irradiated fuel at short cooling times (up to
50-70 years) are the medium half-life fission products 137Cs and 90Sr. Decay heat from thetransuranics such as plutonium, americium and curium only contributes to about 10% of thetotal at 5 years cooling. The transuranic contribution varies only slightly up to cooling timesof 100 years and represents about 50% of the total at a cooling time of about 70 years, atwhich time the short and medium half-life contributors have largely decayed.
Since the capacity of any geological repository is likely to be limited primarily by the
decay heat output, it is beneficial to remove the transuranics by recycling them and also
to separate out and store the 137Cs and 90Sr until they have decayed. Such a strategy of
TRU recycling and Cs/Sr managed storage allows for a considerably increased capacity inthe geological repository and defers the need for follow-on repositories.

1.3 Minor actinide transmutation strategies
1.3.1 Homogeneous/heterogeneous recycling
There are two options
heterogeneous recycling.

forrecyclingminor

actinides,

homogeneousandHomogeneous recycling involves the incorporation of minor actinides as an integral
component of the fuel. In an LWR, this might entail a concentration of a few per cent ofminor actinides dispersed in the fuel rods. Fuel assemblies containing minor actinides

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1. INTRODUCTION

might constitute a fraction or all of the fuel assemblies in the core. Homogeneous
recycling is normally considered best suited to neptunium recycling and less suitable foramericium and curium. This is partly because the burn-out rate of neptunium fits
relatively well with that of conventional uranium or plutonium fuel. Another advantage
is that the radiological shielding requirements for neptunium are not as onerous as those
for americium and curium. Homogeneous recycling has the benefit of allowing the same
fuel design to be used across the core. The disadvantage, however, is that all of the fuelmust be manufactured within a fabrication facility that can meet the shielding andcontainment requirements, which are more demanding than those of conventionaluranium or plutonium fuel fabrication. Neutron emissions and decay heat output fromthe minor actinides are major complicating factors in fuel fabrication. Homogeneous
recycling therefore demands a fuel fabrication plant that has a capacity sufficient to deal
with large fuel throughputs, potentially sufficient to meet the entire refuelling
requirements of the reactor. This could have an adverse impact on the economics ofhomogeneous recycling.
A recent study of the homogeneous recycling of neptunium in PWR MOX fuel has
evaluated the impact on neutron source, gamma source and decay heat [3] and has wellillustrated this impact. The neutron and γ-source characteristics of the spent Pu+Np MOX
fuel (with 11.5 w/o Pu+Np) and the decay heat behaviour of the fuel were comparedagainst the reference MOX case (9.0 w/o Pu). The study demonstrated a slightly reduced
neutron for the Pu+Np case compared with the reference MOX case, while the γ-sourceswere very similar. However, decay heat was slightly elevated in the Pu+Np case.
Heterogeneous recycling is generally considered a better option for americium andcurium. It involves loading the minor actinides in separate, distinct fuel assemblies, often
referred to as “target assemblies” or “target elements”. These can be positioned in thecore in such a way as to optimise burn-out of the minor actinide content, and they may
also be retained for a longer dwell time than the conventional fuel to maximise burn-out.
This increased flexibility in loading is more suitable for americium and curium. In
addition, heterogeneous recycling allows the manufacture of the target assemblies to be
carried out independently of the manufacture of the bulk fuel, and thus the stringent
shielding and containment measures only need to be applied to a relatively small
throughput. The need to ensure compatibility between the bulk fuel and the target fuel,however, is a disadvantage.

1.3.2 Single/multiple recycling
Complete transmutation of the minor actinides is not possible in any reactor system,
meaning that a residual mass of minor actinides will remain at the end of the irradiationof MA-TRU fuel. With a single recycling scheme this residual mass will eventuallycontribute to the inventory of the geological repository. With homogenous recycling, theresidual minor actinide content would be emplaced in geological disposal as an integral
part of the spent fuel. With heterogeneous recycling, the irradiated target assemblies
would be encapsulated for geological disposal, with no further reprocessing required.
Separating minor actinides from fission products and uranium/plutonium is a costly step,
as is incorporating them in homogeneous fuel or heterogeneous targets. With a single
recycling strategy, these very significant costs are only incurred once, though with only alimited reduction of the minor actinide inventory in the repository.
Multiple recycling of homogeneous MA-TRU fuel implies the reprocessing of the bulk
fuel and the separation of the minor actinides for incorporation in a new generation offuels. For heterogeneous minor actinide targets, multiple recycling implies the
reprocessing of the target assemblies and separation of the minor actinides forincorporation in a new generation of target assemblies. This might be accomplished bestby blending the recycled irradiated targets with LWR spent fuel at the beginning of
reprocessing operations. Such blending dilutes the heavier actinides and ensures that thefissile concentration remains high enough for continued cycles of transmutation.

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Multiple recycling has the potential to achieve very low residual levels of radiotoxicity
and decay heat, but suffers from a number of disadvantages:•

The total mass of minor actinides that needs to be separated and processed into
target assemblies is increased, implying higher plant throughputs and higher
overall costs.

•

The total mass of minor actinides that are circulating in the active fuel cycle (i.e.
separation, fabrication storage, transport, in-reactor irradiation) is higher. This
increases the exposure of nuclear workers to radiological dose and increases therisk posed both to workers and the public in the event of a core accident leading touncontained release of radionuclides.

•

The timescales required to actually implement multiple recycling are potentiallyvery long, bearing in mind that each cycle of separation, target fabrication, inreactor irradiation, cooling and transport may take 10 years or more. A
requirement to recycle 10 times, for example, would take at least 100 years, which
is long relative to the lifetimes of nuclear facilities and difficult for decision makers
to commit to.

For these reasons, if it is necessary to implement multiple recycling of MA-TRU fuels,
it is important that the number of recycling processes should be kept to a manageablelevel, for example, no more than 2 or 3.

1.3.3 Storage
An alternative strategy to the recycling of curium is to store the irradiated fuel for >30
years prior to reprocessing. Although the half-life of 242Cm is very short (163 days), it iscontinually replenished because it is from decays of 242Am (242Cm is in secular equilibriumwith 242Am, which undergoes β- decay with a 141 year half-life). 244Cm is the dominantcurium isotope and has an 18.1 year half-life, so that extended storage of >30 years willallow most of it to decay prior to reprocessing.
After reprocessing, the remaining curium could be separated and stored. After asufficiently long period of interim surface storage, the curium and associated decay
products (mainly 240Pu) could be encapsulated and packaged for geological disposal. The
advantage of storage is that the geological repository does not need to accommodate theneutron emissions and heat load of the curium, potentially allowing for an increase incapacity. The disadvantage is that it may be difficult to design a reprocessing plant toseparate curium, and there would of course be a cost associated with the dedicated fuelcycle plants needed to treat and store the curium stream.

1.3.4 Minor actinide recycling in thermal reactors versus fast reactors
Concerns arising from the existence of plutonium and minor actinide stockpiles
accumulated over the years have led to the emergence of systems potentially capable ofreducing these stockpiles through burning and transmutation processes.
A used UOX fuel (UO2 enriched in 235U), originating from a PWR (pressurised water
reactor), contains approximately 1% plutonium (Pu) with 60 to 70% fissile isotopes. After a
few cooling decades, this plutonium represents more than 90% of the radiotoxicity of thefuel. Plutonium multi-recycling allows a significant reduction of the physical quantity ofplutonium while benefitting from the natural energy available.
In a PWR where the thermal neutron flux is enhanced, the production of neutrons is
primarily controlled by the contributions of 239Pu and 241Pu. The even isotopes of
plutonium behave like poisons (σf/σc<<1). With plutonium recycling in a PWR core, twoeffects dominate, each of them reinforcing the other. The deterioration of the isotopic
quality, with the accumulation of the even isotopes, requires an increase in the total
mass of plutonium (via the increase in the plutonium content). The increase in 239Pu and

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1. INTRODUCTION

Pu masses contributes to softening the neutron flux towards the thermal zone where
reduced production rates will decrease the plutonium quality, further increasing theplutonium content needed. In addition, accumulation of minor actinides, in particular of
242
Pu, further contributes to the production of higher actinides (243Am and then curium).
241

In a SFR (sodium fast reactor), the dominant contribution to the neutron balance isprovided by the odd isotopes, although all the isotopes contribute to neutron production
via nuclear fission. Consequently, the production of higher actinides decreases inparticular for minor actinides like americium and curium. In addition, the neutronbalance is more favourable in a SFR spectrum with a surplus of neutrons, which offers
the possibility of 238U conversion and the transmutation of minor actinides.
The deterioration of the plutonium quality in a thermal spectrum makes it possible toconsider limiting MOX (Mixed Oxide fuel PuO2-UO2) multi-recycling in the standard PWR.
The solutions to ease the multi-recycling will come with innovative fuels (MOX-UE,
CORAL, APA) and require, in all cases, a limitation of the initial plutonium content torespect the safety criteria; a contribution from enriched uranium is required to maintainthe burn-up rates objective. These options would make it possible to save, at best,
between 10% and 20% of natural uranium in a nuclear PWR park. Studies to increase theconversion factor (PWR HFC) have been conducted by reducing the moderation ratio andby introducing fertile pins (depleted U) into the core. These changes are limited in termsof natural uranium savings for a PWR nuclear park (at best 30%, i.e. using only 1% of
Uranium ore).
The favourable neutron spectrum in fast neutrons makes it possible to stabilise oreven improve the initial content of plutonium and its isotope content during multirecycling in an SFR core. In contrast to what occurs in a PWR, the improvement and thestabilisation of the plutonium characteristics during SFR multi-recycling will not
deteriorate the core safety coefficients and would thus allow unlimited multi-recycling
with respect to safety. The effectiveness of a fissile core for the reaction chain ischaracterised by the number of neutrons emitted for a core neutron absorption (the ηparameter). The values of this parameter for the 235U (1.9) and 239Pu (2.3) isotopes showthat plutonium is the best choice in SFR cores. The physics of the fast spectrum also
allow the possibility of adapting the SFR core to the production or the burning ofplutonium with appropriate changes in the fertile zones (depleted U) and in the core
design. The production of plutonium (breeding) can be increased by the use of blankets(Breeding Gain: BG = 0.2 for SUPER-PHENIX). This has been demonstrated in the PHENIX
plant where reprocessing has been done since 1980 for the whole core around 5 times,
hence allowing the plant to burn the fuel it has been producing itself (breeding ratio: 1.16).
The consumption of plutonium in SFR can be adjusted so that the plutonium inventory
can either be stabilised or even decreased. Studies show that one can change within thesame plant, the breeding core into a burner core. The CAPRA fuel concept in an EFR corefor instance can burn approximately 75 kg/TWhe of plutonium. A demonstration has
been carried out in PHENIX for some sub-assemblies. Furthermore, the number ofneutrons in excess for achieving the neutron balance can be used to efficiently burnminor actinides being produced by PWR operation and hence reduce to very low levels
the radiotoxicity of nuclear waste accumulated over the years.

1.3.5 Minor actinide recycling in thermal reactors as a technology demonstrator
TRU recycling in thermal reactors could have a role to play as a technology
demonstrator for more advanced minor actinide recycling schemes that might follow
later when, for example, Generation IV systems are operational.
The major benefit of beginning minor actinide recycling in existing thermal reactors
is that it would remove the long-lived radiotoxic nuclides from the fission product waste.
The investment cost for minor actinide recycling will be significant in the fuel cycle

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1. INTRODUCTION

plants in particular. Investment in TRU recycling in LWRs could, however, prepare for
TRU recycling in Generation IV reactors.
If it is possible to achieve sufficient throughputs of minor actinides in LWRs andsufficient burn-out in thermal reactors, then it would be self-justifying.
Investment in the fuel cycle plants accommodating TRU will ease futurecommitments to Generation IV systems, which will be able to fully benefit from such aninvestment. This, in turn, will reassure decision makers, who may otherwise be reluctant
to commit to investment in both fuel cycle plants and Generation IV systems. A relatively
small-scale demonstration programme (perhaps in a limited number of thermal reactors)
would be sufficient to demonstrate the technology on an industrial or sub-industrial
scale before a full deployment, which would have to become economically viable.

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2. POTENTIAL ROLE AND OBJECTIVES OF MINOR ACTINIDE RECYCLE IN THERMAL REACTORS

2. Potential role and objectives of minor actinide recyclingin thermal reactors

This section surveys the various theoretical studies that have been carried out which
examined minor actinide burning in thermal reactors.

2.1 Light water reactor (LWR) studies
In preparation for this Expert Group study, OECD/NEA compiled a list of recent
publications (see Bibliography) relevant to minor actinide recycling in LWRs. Table 2.1
summarises these publications, identifying the country in which the work was carriedout, the study topic and highlighting the main conclusions.
The situation in France is that the nuclear fuel cycle has always been viewed as a very
long-term commitment with a very ambitious target of eventually achieving total selfsufficiency with fast reactors. France is also very keen to recycle minor actinides toreduce long-term radiotoxicity and to this end regards fast reactors as the best means toachieve this end. Although it is accepted in France that minor actinide recycling in LWRs
is technically feasible, any such programme would merely be an intermediate technology
demonstration step towards the eventual goal of minor actinide recycling in fast reactors.
The dominance of the US in the bibliography reflects the particular interest there in
more short/medium term goals which are oriented towards reducing demand in the highlevel waste repository. At the time these studies were carried, the US was interested inanalysing the role of the recycling of irradiated fuel as an alternative to direct disposal.
This strategy would have decreased the capacity utilised in the repository, especially if
the minor actinides are recycled and some heat producing fission products (Sr and Cs)
are removed and stored. The driver in the US at the time was focused on the
short/medium term gains that could be made with TRU recycling in LWRs and this is thereason for the interest shown there. More recently, the US position has been evolving andat the time of writing waste management approaches in the US are being re-examined.
Although the various studies Table 2.1 investigated different scenarios, a number ofcommon themes emerge from the survey of the conclusions:•

There is a strong bias towards PWRs rather than BWRs, the reason for which is notclear. BWR fuel assembly design is more heterogeneous and in particular relies
more heavily on uranium/gadolinium burnable poison rods and for this reason it ispossible that minor actinide recycling in PWRs is considered technically morefeasible. Despite the bias, the fact remains that PWR and BWR neutron spectra are
not so dissimilar and it is reasonable to assume that what is feasible in PWRs is
also feasible in BWRs (though detailed confirmatory analysis would be needed).

•

All of the studies confirm that worthwhile quantities of minor actinides can berecycled in LWRs, with useful quantities destroyed in the irradiation lifetime of thefuel assemblies containing the minor actinides.

•

Many of the studies specifically analysed the impact of minor actinide recycling on
the 3-D core behaviour and were not limited to just lattice calculations of single
assemblies. Many of these studies also analysed the impact on thermal-hydrauliccharacteristics as well. All of the studies that examined these aspects confirmed

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2. POTENTIAL ROLE AND OBJECTIVES OF MINOR ACTINIDE RECYCLE IN THERMAL REACTORS

satisfactory core and thermal-hydraulic behaviour. The fact that so many
independent studies have reached the same conclusion is very strong evidence offeasibility.

2.2 Heavy water reactor (HWR) studies
Atomic Energy of Canada Ltd (AECL) have carried out very extensive studies which
demonstrate the flexibility of CANDU heavy water reactors (HWRs) to burn fuels withdifferent materials compositions. Two recent studies have investigated the potential of
CANDU HWRs for MA-TRU burning [4] [5].
Reference [4] modelled the depletion behaviour of two different types of fuel rods
containing TRU in the CANFLEX advanced fuel assembly bundle. The first fuel type was
MOX fuel comprised of TRU in a conventional UO2 matrix. The second rod type was an
Inert Matrix Fuel (IMF) comprised of TRU in an inert matrix. The MOX fuel assembly was
assumed to reach a very high burn-up of 45 GWd/tHM, while the IMF assembly was taken
to the equivalent energy output. The MOX fuel showed a worthwhile decrease in TRU
inventory of 40% at discharge, while the IMF fuel achieved a reduction of 71%. The longterm heat load for the MOX assembly showed a reduction in the range 30-60%, while thatfor the IMF was 70-80%.
Reference [5] investigated the potential of CANDU HWRs and LWRs for burning
americium. The HWR analysis modelled the irradiation of CANFLEX bundles in a CANDU
reactor, with fuel rods containing various initial fuel compositions of uranium andamericium. The discharge burn-ups of the fuel bundles ranged from 7.5 and 21 GWd/tHM.
In some of the cases examined the americium was assumed to be homogeneously
distributed in the fuel, while in two cases a centre pin consisting of a 3.7 or 7.0 w/oamericium in zirconia was used. The various cases showed transmutation fractions forthe americium ranging from 46% to 79%. These are slightly higher than thetransmutation fractions achieved in the LWR variants examined, the highest of which
was 71%. This indicates that HWRs have the potential to burn minor MA-TRU slightly
more efficiently than PWRs.
Another observation from the HWR study was that HWRs burning americium
recycled from PWR fuel reprocessing in a synergistic fuel cycle can achieve very highsupport ratios. In the synergistic fuel cycle, the americium from reprocessing the spent
fuel from several LWRs is fed to a single HWR. The support ratio is the electrical output ofthese LWRs divided by the electrical output of the americium burning HWR and values ashigh as 20 were obtained: so that 1 GWye of americium burning HWR capacity is needed
to burn the americium from 20 GWye of LWR output. It is important to minimise thecapacity of MA-TRU burning reactors relative to conventional reactors, especially if MATRU burning is less economic, as it minimises the overall economic penalty on the entire
reactor fleet. The equivalent support ratios with MA-TRU burning in PWRs were smaller,
the best ratio being just 3.
The study in [5] shows that a synergistic mix of LWRs and HWRs could be effective in
reducing the inventory of 241Am produced per GWye of useful energy output. One of themain benefits is a reduction in the decay heat output of spent nuclear fuel, which may
ultimately lead to a reduction in the footprint of the geological repository.
A limitation of the studies performed to date on minor actinide burning in HWRs isthat they have been confined to lattice code studies. These are adequate to give an
indication of the transmutation rates that can be achieved with different fuelcompositions, taking into account of the local neutron spectrum. However, additional
analysis will be required to develop the studies further, specifically whole-core nuclear
design assessments. Whole-core analyses are needed to determine the impact on nuclear
safety related parameters such as power peaking factors, reactivity coefficients andshutdown margins, which would be used to confirm safe behaviour of the core under

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normal operation and in fault conditions. During this next level of analysis, it is possiblethat further optimisation of minor actinide loading strategies may be needed to ensuresatisfactory core safety performance.

2.3 High-temperature reactor (HTR) studies
High-temperature reactors (HTRs) have a long history dating to the 1960s. HTRs aregraphite moderated (thermal neutron flux) helium cooled reactors with a ceramic fuelform, capable of very high coolant outlet temperatures (theoretically up to ~1 000°C) andtherefore very high thermal efficiencies. HTR technology demonstrators and prototypes
were built in the 1960s in the United Kingdom (DRAGON), Germany (Jülich HTR, THTR-300)
and US (Peach Bottom, Fort St Vrain), but were abandoned in the mid 1980s partly
because the commercial environment was not yet suitable and partly as a reaction to theChernobyl accident. More recently there has been a revival of interest in HTRs firstly inSouth Africa (where plans to build a prototype pebble bed modular reactor (PBMR) haverecently been abandoned) and in China, where a twin unit PBMR (2x250 MWe) is planned.
HTR cores are known to be very flexible in that they can accommodate a wide rangeof fuel types and the ceramic fuel is capable of very high burn-ups. These characteristics
make HTRs potentially very suited to TRU burning. This was highlighted in a study of
HTRs carried out in the EURATOM 5th Framework HTR-N Project [6]. This examinedseveral different aspects of HTRs, starting with a study of the accuracy of nuclear designmodelling methods and nuclear data libraries when applied to the HTTR-10 and
PROTEUS-HTR research reactor cores. The study also included a cell burn-up benchmark
exercise for an HTR containing plutonium fuel and an assessment of the characteristics
of HTRs as plutonium burners. The HTR-N Project, however, did not carry out any
detailed analysis of HTRs with MA fuels and was intended as a preliminary analysis toprepare for a later MA study.
The follow-up study was eventually carried out under EURATOM 6th Framework. This
was the PUMA Project, which was specifically designed to investigate the potential role of
HTRs in burning minor actinides. PUMA - the acronym stands for “Plutonium and Minor
Actinide Management in Thermal High-Temperature Gas-Cooled Reactors” – was a
Specific Targeted Research Project (STREP) within the EURATOM 6th Framework. The
PUMA project ran from September 1, 2006, until August 31, 2009 and involved a
consortium of 14 European partner organisations and one from the US [7].
PUMA's main objective was to investigate the possibilities for the utilisation andtransmutation of TRU, particularly minor actinides in contemporary and future (hightemperature) gas-cooled reactor designs. The PUMA project assessed the impact of theintroduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance(modelling), reactor (transmutation performance and safety) and reactor/fuel cycle
facility park.
PUMA investigated the core physics of Pu/MA fuel cycles in HTRs and demonstrated
nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating
conditions. The starting point of this investigation comprised the two maincontemporary HTR designs, namely the pebble-bed type HTR, represented by the SouthAfrican PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results
confirm the flexibility of contemporary (and near term future) HTR designs and theirability to accept a variety of Pu- and Pu/MA-based fuels (possibly in combination withthorium). The results show a significant capability for Pu/MA burning, while maintaining
the favourable safety characteristics of conventional uranium-fuelled HTRs.
Pu/MA transmuters are envisaged to operate in a global system with a mix ofdifferent reactor types and fuel cycle facilities. Fuel cycle studies were therefore
performed in PUMA to study synergistic fuel cycles involving HTRs and other reactor
types (LWRs) and fast reactors and to quantify waste streams and radiotoxic inventories.
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2. POTENTIAL ROLE AND OBJECTIVES OF MINOR ACTINIDE RECYCLE IN THERMAL REACTORS

PUMA also assessed the technical, economic, environmental and socio-political impacts.
PUMA showed that a Pu/MA-loaded HTR would give worthwhile reductions on thequantities of TRU in spent fuel and high level waste for geological disposal.

2.4 Survey of experimental studies
A recent IAEA report on minor actinide fuel development [8] summarises theexperimental work that has been carried out worldwide on minor actinide fuels and they
are listed here in Table 2.2. To date, all the experiments carried out on minor actinide
fuels have been at laboratory scale. The irradiations have mostly been in fast reactors,
though many of the results are also relevant to thermal reactors.
Table 2.1: Summary of historical studies of minor actinide recycling in thermal reactors
Study
Preliminary multi-cycle transuranic
actinide partitioning andtransmutation Studies, ORNL/TM2007/24

Can thermal reactor recycling
eliminate the need for multiple
repositories?
C.W. Forsberg, E.D. Collins,
J.P. Renier, C.W. AlexanderMultiple tier fuel cycle studies for
waste transmutation, R.N. Hill etal. ICONE-10-22575

Advanced LWR Multi-Recycling
Concepts, E.A. Hoffman, R.N. Hill,
T.A. Taiwo,
ANS_Trans_Vol_93_Nov05_363364

Year

2007

2004

2002

2005

Country

US

Feasibility of multi-recyling of Pu
and MA in PWRs using combined
non-fertile and UO2 (CONFU) fuel,

2002

2003

Conclusions

Minor actinide (Np-Am-Cm) recyclingin thermal (LWR) and fast spectrum(ABR) systems

LWRs more efficient at burning
MA than ABRs; optimum
performance by irradiating Pu or
Pu/Np in fast spectrum systems
and Am/Cm in thermal
spectrum.

Minor actinide (Np-Am-Cm) recycling
in LWRs

Recycling of TRU increases
Yucca Mountain capacity x 5.
Cs/Sr storage + TRU byfactor 42. TRU recycling in
LWRs and Cs/Sr storage
effective at increasing effective
capacity of repository.

Transuranic recycling in LWRs, LWR
transmutation core, HTR, ADS andfast spectrum reactors

LWRs capable of reducing TRU
waste radiotoxicity to <0.6% of
once-through cycle. Limiting
factor is separation plant losses.

Multiple recycling of transuranics in
LWR in inert matrix fuel (IMF)

Demonstrates that it isneutronically feasible to recycletransuranics in inert matrix fuel(IMF) in LWRs

Transuranic stabilisation in CORAIL
PWR fuel assembly

Confirms that the neutronics of
TRU multiple recycling in PWR
is feasible with the CORAIL
assembly design and variations
thereof.

Transuranic burning in PWR using
fertile free fuel rods

3-D core models used to confirm
feasibility of MA and Pu burningin PWR.

US

US

US

Assessment of transuranics
stabilisation in PWRs
T.K. Kim, J.A. Stillman, T.A. Taiwo,
R.N. Hill, P.J. Finck, M. Salvatores,
PHYSOR 2002

Scope

US

US

E. Shwageraus, P. Hejzlar, M.S.
Kazimi, MIT_CONFU_0164

24

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2. POTENTIAL ROLE AND OBJECTIVES OF MINOR ACTINIDE RECYCLE IN THERMAL REACTORS

Table 2.1: Summary of historical studies of minor actinide recycling in thermal reactors(continued)
French R&D on the Partitioning
and Transmutation of Long-Lived
Radionuclides – An International
Peer Review of the 2005 CEA
Report. OECD/NEA. ISBN 92-6402296-1.

Feasibility of Multi-recycling of Pu
and MA in PWRs using combined
non-fertile and UO2 (CONFU) fuel.
E. Shwageraus, P. Hejzlar, M.S.
Kazimi, Center for Advanced
Nuclear Energy Systems, Nuclear
Engineering Department,
Massachusetts Institute ofTechnology, Cambridge, MA, US
Effect of Thermal-Spectrum
Transmuter Deep Burn-up of
Transuranics on Fast-Spectrum
Transmuter Performance J.A.
Stillman, R.N. Hill
Fuel Cycle Impacts of UraniumPlutonium Co-extraction,
Temitope Taiwo, Frank Szakaly,
Taek-Kyum Kim, Robert Hill
Assessment of a heterogeneous
PWR assembly for plutonium andminor actinide recycling, T.A. Taiwo,*
T.K. Kim, J.A. Stillman, R.N. Hill, M.
Salvatores, P.J. Finck

Actinide transmutation usingpressurised water reactors, M.
Visosky, P. Hejzlar, M. Kazim,
Proceedings of ICAPP ’06 Reno, NV
US, June 4-8, 2006, Paper 6197

2006

2003

2003

2008

2005

2006

Comprehensive review of entire
transuranic fuel cycle by
OECD/NEA International
Review Team (IRT). Fivethermal reactor recycling
strategies considered.

Concludes that although PWRs canminimise continuing production of minor
actinides, they do not have the potential tocompletely burn their own MA production(therefore not good for phase-out
scenarios), Curium accumulates.
Incorporating even small amounts of MAs
has a large impact on the fuel cycle. IRT
views thermal reactors as useful for aninterim plutonium management but not
worthwhile for MA. Needs to be viewed incontext of unified French programme.

Whole-core analysis ofcombined non-fertile and fertile
free fuel pins (CONFU) used torecycle transuranics.

Demonstrates neutronic and T/Hfeasibility. However, does not consider
larger implications for the fuel cycle.

Deep burn-up of TRUs feasible in PWRs.

US

Assessment of transuranic deep
burn concepts in PWR core.
MOX 2x recycling and several
CORAIL variants.

US

Analysis of PWR and ALMR
core performance with different
transuranic fuels including coextraction CX MOX.

Confirms feasibility of re-use of coextraction (CX) recycling material in PWR.

Confirms feasibility of TRU multi- recycling
material in PWR.

US

Multiple recycling of transuranics
in PWRs with CORAIL fuel.
Calculates material
attractiveness.

CONFU-B fuel assemblies demonstrated
multi-recycling of TRU.

US

Analysis of CONFU-B (CONFUburn-down) fuel in PWR,
designed for deep burn oftransuranics.

FranceUS

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2. POTENTIAL ROLE AND OBJECTIVES OF MINOR ACTINIDE RECYCLE IN THERMAL REACTORS

Table 2.1: Summary of historical studies of minor actinide recycling in thermal reactors(continued)
Repository impact of limited
actinide recycling, Roald A.
Wigeland, Theodore H.
Bauer, Robert N. Hill, John A.
Stillman, Proceedings of
GLOBAL 2005,Tsukuba,
Japan, 9-13 October 2005,
Paper No. 496
Preliminary multi-cycletransuranic actinide
partitioning-transmutation
Studies, Emory D Collins etal. Global Nuclear Energy
Partnership Systems Analysis
Working Group, February
2007

26

2005

2007

US

US

Analysis of limited
recycling of transuranics
in PWRs. MOX,CORAIL and inert
matrix fuel (IMF)
examined.

Confirms benefits to repository loadingare possible with limited TRU recycling in
PWRs, just by keeping the TRUs in thefuel cycle. Increase in repository drift
loading is worthwhile but modest withlimited PWR recycling. Fast reactors
needed ultimately to deal with TRU toobtain maximum benefit in drift loading
capacity.

System analyses reporton preliminary
multicycle transuranic
actinide partitioningtransmutation studies

Detailed feasibility study of multiple
partitioning/transmutation cycles in
LWRs.

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2. POTENTIAL ROLE AND OBJECTIVES OF MINOR ACTINIDE RECYCLE IN THERMAL REACTORS

Table 2.2: Summary of minor actinide fuel experiments
Experiment

ReactorFuel materials tested

SUPERFACT-1

PhénixU-Pu, U-Pu-Np, U-Pu-Am, U-Np and U-Np-Am oxides

EFFTRA-T4

HFR Petten

Am in MgAl2O4

X501

EBR-IIAm and Np in metal fuel pins

METAPHIX

PhénixU-Pu-Zr, U-Pu-Zr-MA-Rare Earth, U-Pu-MA metal fuelrods

ECRIX-B and ECRIX-H

PhénixU-Pu-Zr, U-Pu-Zr-MA-Rare Earth, U-Pu-MA metal fuelrods

CAMIX-COCHIX

PhénixAm-Zr-Y oxide, Am-Zr-Y oxide in MgO matrix

FUTURIX-FTA

PhénixU-Pu-Am-Np-Zr metal
Pu-Am-Zr metal
U-Pu-Am-Np nitride
Pu-Am-Zr-N
Pu-Am oxide in Mo cermet matrix
Pu-Am-Zr oxide in Mo cermet matrix
Pu-Am oxide in MgO matrix

FUJI

HFR Petten

Np-Pu-U oxide sphere-pac, Pu-U oxide sphere-pac, Pu-U
oxide vibro-pac, Pu-U oxide pellet

AFC-1

ATR

Pu-Am-Zr, Pu-Am-Np-Zr l, Pu-Zr, U-Pu-Am-Np-Zr, U-PuAm metal

DOVITA

BOR-60U-Pu-Np oxide

HELIOSHFR Petten

Am cermet/cercer

HELIOS AM-1

Joyo

Am MOX and Np-Am MOX

AMBOINE

BOR-60

U-Am oxide, U-Am-Mg oxide

CONFIRM

HFR Petten

Pu-Zr nitride and Am-Zr nitride

GACID

Monju

U-Pu-Am-Np-Cm oxide

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3. ISSUES ASSOCIATED WITH UTILISATION OF MA-TRU FUELS

3. Issues associated with utilisation of MA-TRU fuels

3.1 Recycling modes – homogeneous and heterogeneous
In any type of reactor, there are two possible modes for recycling TRUs and these are
known as homogeneous and heterogeneous recycling:
In the homogeneous recycling mode the TRUs are incorporated as an integral
component of the nuclear fuel. For example, the normal fuel material such as low
enriched UO2 or PuO2/UO2 mixed oxide (MOX) might also contain a minor actinide
component such as NpO or AmO2, smeared homogeneously. Practical considerations
usually demand that the TRU component in homogeneous fuel should be at a low
concentration, representing no more than a few percent of the total heavy metal mass.
Keeping the TRU concentration low limits the radiological dose and decay heat outputs
that need to be managed in manufacturing, transport and handling. It also limits theextent to which the minor actinide loading affects the nuclear design performance of thefuel and core.
Homogeneous recycling has some beneficial features:•

The basic characteristics of the fuel are only modified to a small extent if theminor actinide loading is kept low. This implies that the existing knowledge basefor uranium and MOX fuels is largely still applicable. In particular, if the minor
actinide content is kept low, the underlying fuel performance characteristics
remain close to those of the dominant component of the fuel matrix (either UO2 orMOX), thereby allowing continuity with the vast knowledge base of mainstream
LWR fuel behaviour.

•

Ensuring that all the fuel conforms to the same mechanical, thermal-hydraulicand nuclear design simplifies the design and licensing in all these areas.

On the other hand there are some disadvantages:•

The higher radiological dose and decay heat outputs in manufacturing affect theentire fuel load. The extra costs involved in engineering the fuel manufacturing
plant are imposed on a fuel manufacturing plant that must have sufficient
capacity to accommodate the entire fuel load. Any problems affecting the MA-TRU
fuel would compromise availability of the entire fuel load.

•

The concentration of MA-TRU in the homogeneous fuel is typically quite low andthis may restrict their overall utilisation, if the overall mass of minor actinides
incorporated in the heterogeneous fuel is low.

•

The MA-TRU is constrained to the same core residence time and the same thermal
and fast neutron fluences that normal fuel would be exposed to. Because thefission and capture cross-sections of the minor actinides are relatively low, this
implies that the minor actinides may not be fully transmuted before the fuel isdischarged.

•

Handling, interim storage, transport and disposal of the irradiated fuel iscomplicated by the higher radiological doses and heat outputs associated with the
MA-TRU.

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3. ISSUES ASSOCIATED WITH UTILISATION OF MA-TRU FUELS•

If the irradiated fuel is to be recycled, the mainstream recycling plant must be
capable of handling the elevated TRU content.

The homogeneous option can be regarded as technologically not far removed from
MOX recycling, because much of the existing LWR knowledge base would remain
applicable. It could even be argued that normal MOX fuels already constitute a type ofhomogeneous MA-TRU fuel because of the presence of 241Am from 241Pu decay. Therefore,
it might be possible to regard homogeneous fuels as an extension to existing technology.
This might simplify its qualification, though qualification would still be needed.
Heterogeneous recycling entails loading the MA-TRU in special target fuel rods thatare distinct from the main driver fuel. The MA-TRU target fuel rods will not contribute
fully to neutron multiplication and will therefore need to be supplied with excess
neutrons from the main fuel type. They constitute targets for the excess neutrons andthis is why they are called target rods. The advantages are:•

The design of the MA-TRU target rods, their positioning in the core and theirresidence times can be optimised specifically to maximise TRU utilisation.

•

The manufacturing of the target rods would be carried out in a dedicated plantthat is separate from the normal fuel fabrication plant. The throughput of the MATRU fabrication plant would only need to be a small fraction of the throughput of
the mainstream fuel type, thereby simplifying its design and operation. The
mainstream fuel supply is independent of MA-TRU fuel supply.

The disadvantages are:•Because the major component of minor actinide target rods comprises Np, Am or
Cm, there is very little knowledge base on which to demonstrate satisfactory fuelbehaviour. The knowledge that exists derives from small scale experiments and isinsufficient to guarantee satisfactory performance at commercial scale.

•

Having separate designs for the driver fuel and the target fuel complicates thenuclear and thermal-hydraulic design.

•

Heterogeneous target fuels demand the provision of a second fuel cycle that runs
parallel to the mainstream fuel cycle. All the costs of this second fuel cycle and its
associated facilities add to the costs of the mainstream fuel cycle.

There is no reason why homogeneous and heterogeneous recycling should not be
carried out in parallel, with homogeneous recycling used for Np and heterogeneous for
Am and Cm and there may be benefits from the used of shared facilities.

3.2 Reprocessing
In conventional reprocessing facilities, the minor actinides follow the fission products
into the high level waste stream, eventually being incorporated in vitrified high-level
waste (VHLW). Recycling minor actinides will require modifications to the normal
reprocessing flow sheet to separate them from the fission products and route themseparately or grouped together in some way. While it is relatively straightforward to keepNp with Pu, it is more difficult to separate trivalent Am-Cm from trivalent lanthanide
fission products and even more difficult to separate Am from Cm.
While conventional reprocessing plants produce pure PuO2 as a recycling output
stream, modified flow sheets are now of interest in which the Pu is recovered as a coproduct with U, Np or Am. With >60-70% uranium in the co-product, the proliferationmaterials’ attractiveness is reduced compared with pure PuO2, with higher uranium
fractions giving greater decreases in attractiveness.

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3. ISSUES ASSOCIATED WITH UTILISATION OF MA-TRU FUELS

Irrespective of the flow sheet chosen, the reprocessing plant and associated facilities
will need to provide for the following steps which are additional to those found at current
reprocessing plants:•

provision for chemical separation of the minor actinides from fission products;•

a route for chemically converting the minor actinides to oxide powder that is
subsequently usable in recycled fuel or targets;•provision for shielded handling and storage of the minor actinide powder or coproduct;•

to avoid the need to transport minor actinides or co-products, fuel fabrication
should be co-located and integrated with the separations facilities at the same site.

While the mass of minor actinides is small (about 0.1% of the initial heavy metal
mass of the fuel), the facilities required will nevertheless be substantial and their cost canbe expected to be very significant, both in terms of initial investment and operationalcosts. Unit costs in terms of Monetary Units per kg of minor actinide are likely to be high.
These minor actinide facilities will need to meet the same radiological protection, safety
and environmental impact standards as the conventional reprocessing facilities. A major
goal of fuel cycle R&D will be to maximise the simplicity, flexibility and cost effectivenessof the recycling route.

3.3 Separation technologies
There has been extensive research and development performed on advanced
separation methods over the past decade. The research has been performed across theglobe, including: North America (US, Canada), Europe (EU, France, UK), and Asia (theRussian Federation, Japan, the Republic of Korea, China, India). This research can be
categorised into a few broad categories, the third one of which is most relevant here:•

modifications to the PUREX process;•alternative aqueous processes to recover uranium;•

processes to recover the minor actinides (Am, Cm);•processes to remove fission products for waste management purposes;•

pyrochemical methods of recovering uranium and transuranics.

The PUREX process is being used for commercial spent fuel separations in France, theRussian Federation, Japan and UK and other non-commercial PUREX plants have been
built and operated in other countries such as the US. The PUREX process separates
uranium and plutonium from the fission products and other actinides. Both the uranium
and plutonium streams are very pure suitable for recycling, with only traces of crosscontamination and fission products. However, in recent years sensitivities over the
production of pure plutonium have led to the study of alternative separation schemes
where a co-product of uranium and plutonium and possibly other radionuclides would berecycled instead. Some of these alternative schemes could also be used to separate theminor actinides.
Two primary approaches to this have been developed. In one approach, only uranium
is extracted, and plutonium is left with the other fission products and actinides. This
process is called the Uranium Extraction (UREX) process and was developed in the UnitedStates within the last decade. The other approach is to co-extract uranium andplutonium (as is done in the PUREX process) but to co-strip a fraction of uranium with theplutonium so that pure plutonium is never separated in the process. This process has
been studied as a variation of the UREX process in the US, developed as the COEX™
process in France, and as the NUEX process in the UK.
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Separation of the trans-plutonium actinides from the lanthanides and remaining
fission products for possible transmutation has been a major component of several
national spent nuclear fuel strategies for the middle of the 21st century. This has been thesubject of a substantial amount of research over the past decade in nearly every fuelcycle state.
The UREX approach would require the least initial capital investment and technology
development to implement. The raffinate from the uranium separation process would be
solidified into a form that would provide adequate heat transfer properties. Ideally, thisstorage form would be readily dissolved in nitric acid, for future treatment to recover the
TRUs for future transmutation in reactors. The separation of trivalent TRUs fromlanthanides is a challenging technological problem that has been under intense study forthe past several years. While much progress has been made in the development of
extractants, there does not appear to be a technology currently ready for industrial-scaleuse. The approach of placing the TRUs and lanthanides together in interim storage would
allow for a few additional decades of research and development of new separation
technologies before such technology would need to be implemented.
Two processes that have been extensively developed and tested are theTransuranium Extraction (TRUEX) process and the Diamide Extraction (DIAMEX) process
for separation of the actinides and lanthanides together. Both processes would produce a
relatively pure TRU/lanthanide fraction and the raffinate would contain transition andnoble metals. The primary difference between the processes is the composition of theextractants, with the diamide extractants following the C, H, O, N principle of containing
only carbon, hydrogen, oxygen and nitrogen atoms to allow ease of incinerating spent
solvent and the TRUEX extractant containing phosphorous. Another extractant thatexhibits high separation efficiencies for TRUs and lanthanides is N,N,N’,N’tetraoctyldiglycolamide or TODGA. The development of this extractant is relatively new,
but is under investigation by a number of research laboratories around the world.
The TRUEX process uses a solvent comprised of octyl (phenyl)-N,Ndiisobutylcarbamoylmethylphosphine oxide (CMPO), and tri-n-butyl phosphate in aparaffin hydrocarbon diluent. The TRUEX process is very effective in extracting 3, 4 and 6
valent metals from nitric acid solutions. Complexants, such as oxalic acid, can be added
to reduce the extraction of transistion metals, such as zirconium and molybdenum.
The DIAMEX process originally utilised dimethyl-dibutyl-tetradecylmalonamide(DMDBTDMA) as the extractant in a hydrocarbon diluent. Recently, a new extractant,dimethyl-dioctyl-hexaethoxymalonamide (DMDOHEMA) has been developed thatappears to have better extraction properties than DMDBTDMA. This new extractant has
been tested in a counter-current flowsheet test with actual concentrated high activity
PUREX process raffinates.
Following the separation of the TRU/Ln fraction, an additional process after the
TRUEX or DIAMEX process would be required to separate trivalent actinides fromlanthanides. The state of development of An/Ln partitioning technologies is less than forthe other technologies described above.
Trivalent actinide/lanthanide separation is difficult to accomplish due to the similarities
in the chemical properties of the trivalent actinides and lanthanides. Various solvent
extraction processes have been studied including: the extraction of the lanthanides from thetrivalent actinides with the TALSPEAK process, the coextraction of the trivalent actinides andlanthanides, with selective stripping of the actinides from the lanthanides with the reverse
TALSPEAK process, the di-isodecylphosphoric acid (DIDPA) process, the SETFICS process, thePALADIN process, selective actinide extraction (SANEX) processes using Cyanex 301, the
SANEX-III and SANEX-IV processes (24), as well as processes utilising bis-triazinyl-1,2,4
pyridines (SANEX-BTP). Additional research is being conducted on the GANEX process (groupactinide extraction), which is similar to the combined DIAMEX-SANEX approach, but
attempting to combine them into a single process.

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3.4 Fuel manufacturing
The manufacture of types of nuclear fuel has to be able to meet very stringent
requirements on radiological dose to operational staff. Meeting the acceptedinternational standards on personnel dose uptake is relatively simple for uranium fuels,
but gets more difficult when progressing to U-Pu, U-Pu-Np, U-Pu-Np-Am and TRU fuels
respectively, where TRU denotes the transuranics Np, Pu, Am, Cm and small quantities of
Bk, Cf and Es.
With uranium manufacture it is necessary to control internal dose uptake by limiting
exposure of operating personnel to uranium in air. There is also a requirement to control
surface dose exposure, but the gamma source in uranium fuel is weak and exposure canbe controlled without shielding (usually through control of occupancy time and in somefabrication plants by using automated fabrication and inspection techniques). Neutron
dose rates and radioactive decay heat output are both negligible and do not require any
control measures with uranium fuel. Uranium fuel manufacture remains a largely handson process with the operators being able to handle the fuel pellets and fuel rods without
physical protection.
At a MOX fuel manufacturing plant, the higher internal dose uptake associated with
PuO2 demands that all stages of the manufacturing process leading to the production ofpressure tight fuel rods need to be carried out in a glovebox environment. Moreover, MOX
fuel has higher gamma activity that contributes to external dose and neutron production
and heat output are no longer negligible. The gamma and neutron sources from MOX fuelare dominated by the contribution of 241Am, which builds up following 241Pu decay. The
241
Am present in irradiated fuel is removed when the fuel is reprocessed, but builds up
subsequently as the 241Pu decays. The combination of high 241Pu at the time the fuel isreprocessed and long storage time for the recovered plutonium leads to high 241Am
content and higher gamma and neutron sources. Co-location of separation and fuelfabrication facilities, with close coupling and coordination of separation and fuelfabrication, would minimise the impact of 241Am build-up. In some instances (e.g. US),
long-aged spent fuel, in which most of the 241Pu has already decayed, could be selectively
reprocessed and the subsequent build-up of 241Am almost eliminated.
Stepping up from U-Pu fuel to U-Pu-Np fuel causes a modest increase in gamma andneutron activity of the order of a factor 2 depending on the Np content [9]. This is mainly
caused by increased production of 238Pu from 237Np neutron captures, although there isalso a contribution from the adaptations needed to the plutonium content to maintainreactivity equivalence in the presence of the Np. U-Pu-Np homogeneous fuel assembly
manufacture may therefore be possible at fabrication plants which are similar to thoseused for MOX fuel currently, but with increasing shielding provision for higher Np
contents. Again, close coupling of separations and fuel fabrication facilities would be
effective in reducing the effect of 237Np gamma emissions. It is the in-growth of 233Pa thatis the main contributor to gamma emissions and if fuel fabrication is completed within
about 6 months of separation, gamma emissions from 233Pa can be kept below
problematic levels.
Adding technologically relevant quantities of Am to nuclear fuel (for example in U-PuNp-Am) increases gamma activity, neutron source and decay heat significantly,
approaching a factor of 10 or more [9]. This demands a higher degree of shielding
provision and perhaps the use of remote fabrication and inspection methods to avoid
unacceptable operator doses.
At the top of the hierarchy is TRU fuel, which, in addition to Np, Pu and Am contains
small, but very significant quantities of higher actinides, most importantly Cm. The
gamma activity of TRU fuel can be three times or more higher than U-Pu-Np-Am fuel; theneutron activity is potentially orders of magnitude higher and decay heat two or moretimes higher. With these gamma and neutron activities, TRU fuel manufacture would

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3. ISSUES ASSOCIATED WITH UTILISATION OF MA-TRU FUELS

demand shielding measures comparable with those needed to handle irradiated fuel in
reprocessing plants. All the fuel fabrication and inspection steps would need to be carriedout remotely or automatically. The technological challenges of TRU fuel manufacture
should not be underestimated. Fuel designs that are amenable to remote fabrication,such as vibro-compaction, are clearly well suited to TRU fuel if they can be demonstrated
to perform satisfactorily.

3.5 Fresh fuel transport and handlingUO2 and MOX fuel assemblies for LWRs are transported in unshielded containers.
With the increased gamma and neutron doses from U-Pu-Np, U-Pu-Np-Am and TRU fuels,
there will be a requirement to install adequate shielding in or around the transport
packages and also to ensure the handling of fuel assemblies during export from thefabrication plant and upon receipt at the LWR plant is protected with sufficient shielding.
A potential issue with fuel handling is that of transferring heated assemblies into wet
storage. Since the heat output from minor actinide fuel is likely to be much higher thanUO2 and MOX fuels, a cooling phase may be required before minor actinide fuel can betransferred. There is also a potential question as to whether dry storage would be feasible.

3.6 Irradiation testing
Prior to commercial scale deployment of homogeneous or heterogeneous MA-TRU
fuel there will need to be a programme of materials properties measurements andextensive irradiation testing. Materials properties measurements would initially be madeon un-irradiated samples.
Irradiation testing would start with small-scale irradiation tests of short length fuelrods typically done in facilities such as the OECD/NEA Halden Reactor Project (HRP).
Irradiated samples would typically be subjected to post-irradiation examination (PIE) to
confirm satisfactory irradiation performance and provide irradiated fuel properties
measurements.
Following small scale irradiation tests there would be a requirement to carry out
demonstration irradiations in a commercial reactor, which might perhaps involve theirradiation of small numbers of fuel rods or a small number of complete assemblies.
Typically, fuel rods might be discharged after 1, 2 and 3 irradiation cycles to confirm theirradiation behaviour is as expected. The objective would be to demonstrate satisfactory
irradiation endurance prior to loading minor actinide assemblies at commercial scale.
The entire process from initial design to start of commercial scale loading might be
expected to take at least 10 years and therefore irradiation testing is a rate limiting step.
Commercial scale loading also pre-supposes the availability of a suitable fuel fabrication
plant, of which none exist at present. The overall timescale required is therefore rather
longer, at least 15 to 20 years, so that MA-TRU fuels need to be seen as a medium-term
option that will definitely not be available in the immediate future.
The degree of rigour required of the irradiation testing and demonstration
programme depends on the type of fuel selected. It could be argued that for U-Pu-Np fuelmuch of the existing fuel performance database for MOX fuel will remain valid andtherefore a less demanding test programme would be required, facilitating early
introduction. However, for TRU fuel, particularly a novel fuel type such as TRU fuel in
non-uranic matrix [i.e. inert matrix fuel (IMF)], a more rigorous and presumably longer
testing programme would be required.

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3. ISSUES ASSOCIATED WITH UTILISATION OF MA-TRU FUELSWith any new fuel design there is a degree of technical risk and the possibility of poor
in-core behaviour becoming manifest only after the start of commercial-scale loading.
Utilities would be understandably reluctant to bear this risk alone and some degree ofgovernmental underpinning may be needed to encourage utilities to adopt minor actinide
burning. In many respects, the nature of the benefits of minor actinide burning would besocietal and not specific to the utility itself. In this context, governmental underwriting oftechnical risks would be at the very least desirable and might even be essential.

3.7 Licensing
One of the advantages of burning minor actinides in LWRs is that it would be possible
to benefit from the comprehensive licensing regimes that are already in place. Even witha very radical TRU fuel, it would be a sensible objective to try to retain intact as much of
the existing licensing basis as possible. With a small loading of minor actinides in thecore it might be possible to keep within the existing nuclear design and thermalhydraulic limits so that the overall safety behaviour of the fuel and core fits withincurrent limitations. In this case, the only alterations to the safety case required would be
those which specifically relate to the fuel performance of the fuel assemblies that containthe minor actinides. This might have a carry-over to the rest of the safety case, forexample, if fuel failure source terms are affected, but the overall expectation is one
where the safety case is largely unaffected.
For higher minor actinide loadings it is probable that the existing nuclear design andthermal-hydraulic limits will be exceeded, in which case there will be an impact on the
broader safety case. There may be a need for core modifications such as control rod andburnable poison design changes or other core modifications to adjust the reactivity
feedback coefficients.

3.8 Irradiated fuelWith homogeneous MA-TRU fuel, there will be relatively minor implications forirradiated fuel arisings. The presence of minor actinides may reduce the mean discharge
burn-ups if it is not possible to compensate with higher initial fissile loading, which will
give a corresponding increase in irradiated fuel discharges. The inventory of theirradiated fuel will be altered by the minor actinides, with higher specific heat loadings
and neutron outputs the likely outcome.
With the inhomogeneous strategy there would be several major effects on irradiatedfuel arisings: the first is that there would be two distinct types of fuel discharged, theconventional driver fuel assemblies and the minor actinide target assemblies, possibly
necessitating some duplication of back-end fuel cycle facilities. The driver fuel would beentirely conventional, though there would possibly be differences arising from the higher
initial fissile loading required and also possibly a higher mass discharged if the meandischarge burn-up cannot be kept constant. The minor actinide target fuel would becompletely different in nature to the driver fuel, possibly with a higher decay heat output
and higher neutron activities. The target fuel might require a different regime for in-pondcooling and different arrangements for transport. Whether the target fuel is intended forgeological disposal or reprocessing and recycling, there will be completely different
requirements to those of the driver fuel.

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3. ISSUES ASSOCIATED WITH UTILISATION OF MA-TRU FUELS

3.9 Waste
With homogeneous minor actinide recycling the waste arisings will hardly be affectedcompared with a conventional fuel cycle. If the irradiated fuel is reprocessed, there will
be a higher inventory of TRUs compared with more conventional fuels. If the minor
actinides are recycled further, then there would be little or no impact on theincorporation rate of fission products in the VHLW glass. However, if the TRUs are notrecycled further, there is the potential for the higher neutron source to deleteriously
affect the vitrification incorporation rate. The effect of this might be to increase thenumber of VHLW glass containers per GWye. To determine whether this actually turns
out to be the case would require a detailed mass-flow scenario analysis of the entire fuelcycle, accounting for the lower mass flow of TRUs in the first generation of fuelreprocessed (i.e. the minor actinides from this fuel are recycled). Although recycling of Np
and Am destroys a high fraction, the Cm continues to build up and the neutron sourcefrom 244Cm is potentially a limiting factor.
With heterogeneous recycling, the situation regarding VHLW waste volumes ischanged significantly, in that there are now two fuel types that may potentially be
reprocessed: the driver fuel and the minor actinide target fuel. If only the driver fuel isreprocessed, then the VHLW waste volume is likely to be reduced, because thetransuranic mass in the VHLW will be lower if some is recycled in target assemblies.
However, if the target assemblies are also reprocessed, these will have a very highneutron activities and decay heat outputs and unless some of the transuranics arerecycled further, there is the possibility that the vitrification incorporation rate may be
adversely affected. Again, detailed mass-flow scenario analysis is needed to provide
definitive answers.

3.10 Impact on geological disposal
Reference [4] shows that americium burning in both LWRs and HWRs leads tosignificant reductions in the decay heat load in the geological disposal facility. This isbeneficial because the decay heat load generally limits the capacity of a geologicaldisposal site.

3.11 Overall timescales
Bearing in mind, as discussed in Section 3.5, the need for extensive radiation testing,the earliest possible date for commercial scale minor actinide recycling in LWRs would be
about 2020, with perhaps 2025 or 2030 being more realistic. If minor actinide recycling
becomes established, the subsequent timescales for worthwhile recycling programmes
will be measured in decades and it is quite probable that minor actinide recycling
instigated in new build LWRs between 2020 and 2030 would still be active when theseplants reach the ends of their lifetimes in 2080 or 2090.
While the overall timescale will depend on the specific scenario and in particularwhether there is multi-recycling of the minor actinides, the fact is that any minor
actinide recycling programme is likely to require a very long-term commitment of many
decades. This does not fit well with the time horizons of commercial utilities and isanother reason why governmental involvement will be necessary to provide thenecessary incentives.

3.12 Non-technical considerations
As already hinted at in this section, non-technical issues that will need to beaddressed include economics and societal: as will be discussed in Chapter 7, there will besignificant costs associated with minor actinide recycling and the benefits are mostly in

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3. ISSUES ASSOCIATED WITH UTILISATION OF MA-TRU FUELS

favour of broader society and not for the utilities. Therefore the question of economics is
not just an issue for utilities and fuel cycle companies, but also for society andgovernments. Mechanisms will need to be found by which the wider societal benefits will
be recognised and utilities compensated for the additional costs they will incur.
It is possible that in some countries societal concerns may preclude the continued use
and further expansion of nuclear energy unless there is a definite resolution of spent fueland waste disposal strategies. In some cases, such strategies may demand some form ofminor actinide recycling.

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4. FUEL CYCLE ISSUES

4. Fuel cycle issues

4.1 Source terms
This section discusses the source terms from MA-TRU fuels for heat output, gamma
emissions, neutron emissions and radiological doses in the context of a specific example.
A publication from the bibliography [9] provides source term data that is useful here forillustration. It was chosen as an example because it represents a very detailed andcomprehensive study and includes all the elements needed here to illustrate thepracticalities on transuranics recycling in LWRs. The Taiwo paper describes an analysis ofmulti-recycling scenarios in a PWR using the CORAIL assembly concept that originated in
France [15-17]. This was one of a number of assembly designs that were developed in
France initially in the context of plutonium recycling, but which were also investigated in
the context of minor actinide recycling [18-21].
CORAIL (Figure 4.1) is a PWR fuel assembly that is dimensionally compatible withconventional assembly designs, but differs in that it has a heterogeneous distribution offuel rods. The central part contains conventional low-enriched UO2 fuel rods which act as
the driver. Around the periphery is a zone with fuel rods containing Pu. Pu-Np, Pu-Np-Am
or TRU. The CORAIL fuel assembly can be categorised as a homogeneous recycling option,with the transuranics concentrated in the peripheral fuel rods. The CORAIL assembly has
84 MOX fuel rods out of the 264 total, so that the TRU rods represent one third of the fuelmass.
Figure 4.1: CORAIL PWR fuel assembly

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4. FUEL CYCLE ISSUES

The main objective of recycling the transuranics in the scenario studies in is not somuch to transmute them as to store them in active circulation in the core and thereby
delay them going to the geological repository (this is referred to as the “delay line”
concept). Recognising that the transuranic inventory in active circulation increases ineach successive recycling, a possible end solution could be transmutation in a fast reactor
system. This is the same end-point as envisaged in France, but the delay line strategy
allows the fast reactors to be introduced at a much later stage.
Table 4.1 shows equilibrium isotopic composition data extracted from the Taiwo
paper [9] for four scenarios:•

equilibrium recycling of Pu only;•

equilibrium recycling of Pu+Np;•

equilibrium recycling of Pu+Np+Am;•

3rd recycling of all transuranics.

Table 4.1: Equilibrium isotopic compositions for multi-recycling scenarios from [9]Pu Equil

Pu-Np Equil

Pu-Np-AmEquil

TRU 3rdrecycling

Cooling time (y)

5

5

5

5

Uranium enrichment (weight %)

4.62

4.9

5.5

4.98

TRU content (%)

8.45

9.93

16.17

8.39

Fissile content (%)

44.67

43.22

36.23

45.23

237Np (%)

1.66E-3

2.84

2.23

3.00

239Np(%)

0

0

0

0

238Pu (%)

3.55

6.13

10.95

5.34

239Pu (%)

34.2

33.25

28.27

34.17

240Pu (%)

23.33

22.38

20.55

24.37

241Pu (%)

10.48

9.98

7.89

9.84

242Pu (%)

27.4

24.44

18.99

11.50

241Am (%)

1.04

0.99

6.01

5.63

242mAm (%)

-

-

0.06

0.03

243Am (%)

-

-

5.05

3.28

242Cm (%)

-

-

-

0

243Cm (%)

-

-

-

0.03

244Cm (%)

-

-

-

2.22

245Cm (%)

-

-

-

0.43

246Cm (%)

-

-

-

0.07

Target material

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4. FUEL CYCLE ISSUES

Equilibrium means that the materials are recycled many times until the inventories
no longer change between successive recycling. In the case of full transuranic recycling,equilibrium is only reached after a very large number of recycling that has no practicalrelevance and therefore data only for the 3rd recycling step is shown here, which is point
judged to be the practical limit on account of the high neutron source.
In each case the cooling time between discharge and the start of the next irradiation
cycle is 5 years. Although the compositions are only valid in the specific scenario
considered in the Taiwo paper, they are nevertheless useful here as representative
compositions which yield illustrative source terms. The Taiwo paper assumes anequilibrium discharge burn-up of 45 GWd/t in a 3-batch fuel cycle. The 235U enrichment isadjusted as necessary to maintain reactivity over the required discharge burn-up.
Table 4.2 and Figure 4.2 show the heat, gamma and neutron source terms for the four
isotopic compositions at the time of fuel fabrication. Decay heat ranges over a factor of ~5and gamma output ranges over a factor ~14. However, the neutron source increases bymore than two orders of magnitude for full TRU cycle (note the logarithmic vertical axis
in Figure 4.1). While the step from conventional Pu fuel to Pu-Np is only a factor of ~2 forheat and gamma and smaller for neutrons, there is a much larger increase in the neutron
source for Pu-Np-Am and full TRU recycling, because of the strong neutron source from244
Cm. Continuing full TRU recycling beyond the 3rd recycling step causes the neutron
source to increase even further and one of the reasons cited for stopping at 3rd recyclingis that at this point the neutron source is considered as strong as might be practicallymanageable.
Table 4.2: Decay heat, gamma and neutron sources during fuel fabrication for different
TRU isotopic compositions
Pu
Equil

Pu-Np Equil

Pu-Np-AmEquil

TRU 3rdrecycling

Decay heat (W/tHM)

1960

3766

11600

9227

Gamma output (W/tHM)

0.57

1.08

7.79

5.58

Neutron output (n/s/tHM)

1.20E+08

1.84E+08

4.60E+08

2.43E10Target material

Figure 4.2: Decay heat, gamma and neutron sources during fuel fabricationfor equilibrium isotopic compositions
1.0E+12
1.0E+11

Heat (W/tHM)
Gamma output (W/tHM)

1.0E+10

Neutron source (n/s/tHM)

1.0E+09
1.0E+08
1.0E+07
1.0E+06
1.0E+05
1.0E+04
1.0E+03
1.0E+02
1.0E+01
1.0E+00
1.0E-01
Pu

Pu-Np

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4. FUEL CYCLE ISSUES

4.2 Fuel fabrication
The Taiwo paper evaluates surface and 1 m doses for the CORAIL-Pu, CORAIL- Pu+Np
and CORAIL-TRU fuel assemblies and these are listed in Table 4.3. In each instance,
gamma activity dominates the dose rate, with the contribution from neutrons only
significant for CORAIL-TRU (the neutron doses from 2nd recycling TRU are orders ofmagnitude higher).
Table 4.3: Dose rates (mSv/h)
Dose rates (mSv/h)

CORAIL-Pu at equilibrium

CORAIL-Pu+Np atequilibrium

CORAIL-TRU

Total dose at pellet surface

49.0

88.0

151.1

Total dose at fuel clad
surface

1.3

1.9

44.6

Total dose at 1 m

4.2E-3

5.3E-30.2

2nd recycling

Compared with CORAIL-Pu, there is only a modest increase in dose rates for CORAILPu+Np, suggesting that with suitable design and operation modifications, existing MOX
fuel fabrication plants may be adapted for CORAIL-Pu+Np. However, there is a large
increase in dose rates for CORAIL TRU, especially at the fuel clad surface and the 1 mdose, even in the 2nd recycling step.
With a low throughput of Pu-Np, the increase in occupational exposure with CORAILPu+Np might be manageable because the increased exposure would apply only for asmall proportion of the time. However, with a high throughput of Pu-Np, it might benecessary to install additional gamma shielding and/or increase the level of automationin fabrication and inspection operations. For CORAIL Pu-Np-Am fuel the gamma source
still dominates dose rates, but is now substantially higher and this would demand adifferent type of fuel fabrication plant, with significantly more gamma shielding and/or ahigher degree of automation of the fabrication and inspection operations.
Conventional MOX fuel fabrication operations can generate significant quantities ofrecycled material from fuel pellet grinding operations and from defective pellets or
defective fuel rods. Recycling operations will be more complicated for the morechallenging fuel types. Co-location of the fuel fabrication plant with the reprocessing
plant would simplify the recovery and recycling of such material, especially for Pu-NpAm and TRU fuels and might avoid the need to construct a separate recycling facility
alongside the fabrication plant.

4.3 Transport
To date, minor actinide fuel rods have only been manufactured on very small scales
suitable for laboratory and irradiation testing and there has been no requirement totransport minor actinide fuels in commercially relevant quantities. Strategies will need to
be developed for transporting fresh minor actinide fuel assemblies to power plants andfor transporting the spent fuel assemblies. The unshielded fresh fuel transport containers
used for UO2 and MOX will not be suitable for minor actinide fuels and shielded transport
containers will be required. One approach may be to develop specifically designed
shielded transport packages. Another possibility might be to adapt existing spent fueltransport packages for this purpose. Both cases will have implications for at-reactor fuelreceipt and storage operations, with implications for handling requirements(e.g. crane/elevator load requirements), radiological dose uptake and accident scenarios(e.g. dropped fuel accidents).

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It is likely that existing spent fuel transport packages may be useable as-is oradaptable for transporting irradiated minor actinide fuels. However, the higher heat
output and neutron output may require special strategies to be developed for loading theminor actinide fuels, perhaps as part of shipment of normal UO2 fuel assemblies, which
could be used to partially shield the elevated neutron emissions and dilute the heat load.
For both fresh and spent minor actinide fuel transport there is no immediate prospectof a commercial requirement developing and until this happens there will be no
incentive for fuel transport companies to make the necessary investments.

4.4 Utilisation rates
Utilisation rates for the minor actinides will depend on the intended mission of minor
actinide burning. A minimum goal might be to ensure that the future equilibrium
inventory of minor actinides should be stabilised, implying that the utilisation rateshould match the rate and which the minor actinides accumulate in irradiated fuel. A
more ambitious goal might be to recycle minor actinides from historic irradiated fuelstocks, which would require a higher utilisation rate. The minor actinide core loadings
required for both scenarios would depend also on whether the MA-TRU fuels areirradiated just once or multiple recycled to equilibrium. In the latter case, the minor
actinide mass flows from equilibrium and historic irradiated fuel will be supplemented
by the mass flows from recycled minor actinide targets.
France has already decided that LWRs are not the most suited to the more ambitious
goal of reducing the historic minor actinide inventory and prefer to rely on fast reactors
for this task. Indeed, in France, only part of the spent fuel has been reprocessed (around10%). The plutonium from the reprocessed fuel has been used to fabricate mixed oxide
fuel (MOX) for use in LWRs. This action did enable a reduction in the rate of build-up of
plutonium and also the number of spent fuel assemblies in repositories (almost 1/5 UOXspent fuel sub-assemblies). Some MOX spent fuel sub-assemblies have been reprocessedbut only for demonstrating the ability of current reprocessing technology. The minor
actinide inventory of the reprocessed fuel has been incorporated in VHLW glass and ispractically irrecoverable for further transmutation. In the US the situation is different
with the historic irradiated fuel still in interim storage and therefore the historic minor
actinide inventory remains accessible. The US position was different in that the main
goal was the extension of the effective lifetimes of the geological repository and a rangeof different utilisation rates could potentially have satisfied this requirement. As noted
earlier, however, the US position is now evolving and the goal may be modified.
It is helpful to quantify the minor actinide inventories as a fraction of the total
plutonium content:For PWR UOX fuel irradiated to 45 GWd/t the ratios are:•Np/Pu ~3%;•

Am/Pu ~7%;•

Cm/Pu ~0.4%.

With the CORAIL assemblies, the maximum possible TRU fuel fabrication
requirement is about 6-7 tHM/GWye at a burn-up of 45 GWd/t (ie 30% of the total fuelrequirement of ~22 tHM/GWye). Reference [9], which assumes all the TRU is continuously
recycled, indicates a total TRU content of ~7.5% for 2nd recycling CORAIL, of which 1.76%
is 237Np, so the maximum mass of Np recycled per year is ~9 kg/GWye. The Am content is
7.7% of the total, equating to ~38 kg/GWye.
This study does not take into account the impact of such recycling on the safety ofthe plant. In practice, after 2 or 3 recycling processes, the plutonium content needs to be
increased to recover the necessary criticality margins for operating the plant and hits the
12% equivalent plutonium content limit that is acceptable for safety reasons. The impact

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4. FUEL CYCLE ISSUES

of minor actinides on the safety criteria is even larger than the plutonium with roughly
1% of minor actinides being equivalent to 3% of plutonium.

4.5 Irradiated fuel inventories
Assuming that the MA-TRU fuel has the same mean discharge burn-up as
conventional fuel, then the total inventory of irradiated fuel will be unaffected. However,
the activity of the minor actinide component will be much higher and this implies
modifications to the subsequent storage, transport and reprocessing operations.

4.6 Radiotoxicity and environmental impact
Figure 4.3, which is reproduced from [9], shows the evolution of radiotoxicity withdecay time for seven different fuel types, ranging from conventional UO2 fuel, to six
different variants of plutonium and transuranic recycling in the CORAIL fuel assembly(two of which do not recycle plutonium have not been considered here).
Compared with the once-through UO2 fuel cycle, all the CORAIL recycling options are
effective in reducing the radiotoxicity of spent nuclear fuel at cooling times of >100 years.
Although all the CORAIL recycling options result in an increase in radiotoxicity at shorter
cooling times, this is not a concern because the spent fuel emplacement would not takeplace then.
The CORAIL-Pu and CORAIL-Pu+Np options have virtually identical equilibrium
radiotoxicity profiles, a factor 2-3 lower for cooling times between 100 and 1000 years andsomewhat higher for very long cooling times. A much larger benefit is obtained with theequilibrium CORAIL-Pu+Am fuel cycle, especially for cooling times of a few hundred to1000 years.
The largest benefit is obtained in the equilibrium CORAIL-TRU fuel cycle, which
reduces the radiotoxicity of spent fuel by about two orders of magnitude compared withthe once-through LWR fuel cycle at cooling times >500 years. However, the neutron andgamma emissions are so strong that such a fuel cycle would probably be impractical to
implement. Limiting CORAIL-TRU to 2nd or 3rd recycling would be more realistic, but will
yield a smaller benefit in radiotoxicity.
Figure 4.3: Radiotoxicity versus decay time for CORAIL fuel variants (based on [9])44

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5. FUEL AND CORE DESIGN

5. Fuel and core design

5.1 Fuel design
The impact of minor actinide recycling on fuel design depends on whether
homogeneous or heterogeneous recycling is adopted:
With homogeneous recycling it is feasible for the fuel assembly design to remain
essentially unchanged from that used for the conventional UO2 or MOX fuel. Any
modifications that are required to the fuel rod design are likely to be relatively minor
ones to accommodate higher internal gas pressures from helium and fission gas release,such as decreased stack height to accommodate an increase in plenum volume.
Maintaining the standard fuel assembly mechanical and thermal-hydraulic designs
ensures full compatibility with existing fuel. Because the presence of minor actinides in
the fuel tends to decrease reactivity, it may be necessary to increase the initial fissile
loading of 235U or plutonium. If LEU is the main fissile driver material, there is the issue ofthe 5.0 w/o fabrication limit to consider. Currently, commercial LWR fuel fabrication
plants are only permitted to manufacture LEU fuel with a 235U enrichment below 5.0 w/o,
for criticality safety reasons. Since many LWRs are already approaching this limit, therequirement to maintain reactivity may cause it to become the limiting factor even
earlier. However, fuel fabricators are beginning to consider how they may be able toextend the criticality safety limit and it is possible that the limit might be relaxed by thetime commercial LWR utilities are ready to recycle minor actinides.
With heterogeneous recycling, the minor actinide bearing fuel rods will necessarily be
different to those of the driver fuel. For example, in the CORAIL assembly the TRUs areloaded in the peripheral fuel rods and these are distinct to the conventional LEU rods ofthe driver fuel. The LEU and TRU rods are identical in terms of external dimensions, but
the TRU rods may have different internal designs to accommodate additional helium andfission gas release. Other approaches may go further and use a different mechanicaldesign for the TRU rods, introducing the complication of demonstrating mechanical andthermal-hydraulic compatibility. It is even conceivable that the entire fuel assembly
design may be non-standard, but this would further exacerbate the compatibility issues.

5.2 Core design
A recent study has determined the impact of loading minor actinide target rods onthe core design of a 17x17 PWR core [10]. The objective was to determine the impact ofloading minor actinide target rods containing americium and curium in roughly the
proportion they arise from irradiated PWR fuel after a cooling time of 6 years. To this end,
the exact design of the target rods was not of interest and the entire focus was on theimpact they would have on the nuclear design parameters. The starting point was togenerate an equilibrium UO2 core in order to define a reference case for comparison:

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5.2.1 Equilibrium UO2 core
A 4-loop PWR with 193 17x17 fuel assemblies was used for the reference core. The
targeted energy output was an industry standard of 18 months. The reference
equilibrium core design, calculated using a 3-D nodal core design code, has a cycle burnup of 18.3 GWd/t, which corresponds to 477 effective fuel power days (efpd) at thenominal full power rating of 38.4 MW/tHM. The load factor corresponds to 87%. The
84 feed assemblies have initial enrichments of either 4.38 or 3.98w/o, with an overallmean of about 4.2 w/o.
Gadolinia (Gd2O3) admixed with the UO2 was used as a burnable absorber to (i) reduce
the excess reactivity early on in the fuel cycle and (ii) to keep the power distribution
within the design limits. As is standard practice, the Gd2O3-doped rods were enriched to alower concentration (2.0 w/o 235U) than the bulk of the assemblies to counteract the lower
thermal conductivity of Gd2O3-doped fuel. The remainder of the fuel was enriched to aslightly higher than nominal value to give the required mean 235U enrichment.

5.2.2 Equilibrium Am-Cm core
For the purposes of this study, it was decided to adopt the simple approach of loading
the Am-Cm target fuel rods in the control rod guide tubes of the standard PWR fuelassembly. It is acknowledged that such an approach may cause practical engineering
concerns, such as maintaining adequate coolant flow in the restricted flow channel
between the target rods and the control rod guide tubes. However, for the purpose ofdetermining the impact on the nuclear design, such concerns were set to one side in the
expectation that a more considered engineering approach would give similar results forthe core behaviour.
Recognising the need to restrict the total Am-Cm loading, only one third of fuelassemblies were assumed to contain Am-Cm rods and each of those assemblies that was
assumed to contain 24 rods in the control rod guide tubes. Obviously, the Am-Cm bearing
fuel assemblies must avoid the control rod locations and this helps to minimise theimpact on control rod reactivity worths.
The target rods were assumed to consist of AmCmO2/CmO2 in an inert matrix andclad in Zircaloy-4. The inert matrix was of no interest to this study and assumed to be
neutronically transparent. The remainder of the fuel rods in the Am-Cm assemblies was
assumed to contain 4.2 w/o 235UO2 (with no Gd2O3 dopant). The Am-Cm concentration was
adjusted so that the variation of reactivity with burn-up was similar to that of a fuelassembly containing discrete B-10 poison rods. The objective here was to take advantageof the Am-Cm to perform as a burnable poison and help control overall reactivity.
To match approximately the reactivity of a 4.2 w/o UO2 fuel assembly doped B-10
burnable absorbers it was found that a loading of 4.2 kg of Am-Cm was required,distributed equally among the 24 guide tubes of the assembly. In every 84 assembly
reload, 24 of the assemblies were assumed to contain 24 Am-Cm rods. The total mass of
Am-Cm loaded in every 18 month cycle is therefore close to 100 kgHM.
Table 5.1 shows the isotopic composition of the Am-Cm feed to the minor actinide
target rods. This corresponds to the isotopic composition obtained from PWR fueldischarged with a burn-up of ~50 GWd/t and cooled for 6 years.

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5. FUEL AND CORE DESIGN

Table 5.1: Isotopic composition of Am-Cm feed mix
Nuclidew/o

Am-241

63.85

Am-242<0.01

Am-24325.35

Cm-242<0.01

Cm-2430.09

Cm-244

9.78

Cm-2450.82

Cm-246

0.11

The Am-Cm rods were assumed to reside in the core for three 18 month cycles, tomaximise their overall burn-up (which is higher than that of the UO2 fuel, roughly
equivalent to 60 GWd/t, though in the inert matrix the actual burn-up expressed per
tonne of heavy metal is much higher). The cycle length of the equilibrium Am-Cm-doped
core was 17.1 GWd/t, which is a reduction of 1.2 GWd/t on the reference UO2 core, eventhough the 235U loadings are very nearly equivalent. This represents loss of approximately
31 efpd, due to the negative neutron balance of Am-Cm, which has important economic
implications for an utility. This could be avoided by increasing the initial 235U enrichment,
but this still means that the utility incurs an unavoidable cost penalty.
The balance between minor Am-Cm destruction and production in the core is quitefavourable. In the Am-Cm rods the equilibrium core destroys approximately 70 kg of AmCm in every 18 month cycle, from the 100 kg loaded initially. In the remainder of the core,
which contains UO2 fuel rods, approximately 35 kg of Am-Cm are produced. The net AmCm destruction is therefore ~-35 kg Am-Cm per 18 month fuel cycle. A single reactor
loaded with Am-Cm in this way is able to destroy all the Am-Cm it generates itself andthat of a matching reactor loaded entirely with UO2 fuel. Although Am is destroyed very
effectively, in common with other reactor systems, the Cm tends to accumulate throughneutron captures.

5.2.3 Nuclear design parameters
Table 5.2 indicates how the nuclear design limits are modified in the equilibrium corewith Am-Cm target rods compared with the reference UO2 equilibrium core.

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5. FUEL AND CORE DESIGN

Table 5.2: Impact of Am-Cm target loading on equilibrium nuclear design parameters
Parameter

Am-Cm impact

SignificanceRadial power peaking
factor F∆H

Marginal violation of nuclear
design limit

Probable mitigation with coreloading pattern optimisation
and/or relaxation of limits

Axial power peaking
factor FZ

Small

None

Overall power peaking
factor FQ

Small increase

Probable mitigation with coreloading pattern optimisation
and/or relaxation of limits

Moderator temperature
coefficients

Marginal violation of nuclear
design limit

Probable mitigation with coreloading pattern optimisation
and/or relaxation of limits

Doppler coefficient

Small increase

None

Boron coefficients

Slight reduction in magnitude(less negative), marginal
violation of nuclear design
limits

Probable mitigation with coreloading pattern optimisation
and/or relaxation of limits

Control rod reactivity
worthsSlight increase in magnitude(more negative) within
nuclear design limits due topower density re-distribution
effects

None

Shutdown margins

Small increase in end-ofcycle shut-down margin

None

Delayed neutron fraction

Small

None

Boration limits

Moderate impact on boron
concentrations within normal
nuclear design limits
envelope

None

The main observations are:48•

The radial power peaking factor F∆H is increased (see Table 5.3), slightly above thenuclear design limit of 1.528, but the study just used the equilibrium UO2 assembly
loading pattern without re-optimisation. It is judged that the slight violation could
be avoided with loading pattern optimisation.

•

The overall power peaking factor FQ is increased slightly, in line with the higher F∆H,
though within the nuclear design limits.

•

The moderator temperature coefficients are more negative with Am-Cm loading.
The precise effect varies depending on the core condition, but the incremental
changes are all modest, with the magnitude of the moderator coefficient changing
by no more than 8 pcm/°C. This may be partly caused by the reduction in overallmoderator/fuel ratio as a consequence of loading the Am-Cm target rods in thecontrol rod tube locations. Tables 5.4 and 5.5 indicate the Hot Full Power (HFP) and
Hot Zero Power (HZP) moderator temperature coefficients for the two equilibrium

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5. FUEL AND CORE DESIGN

cores. The calculations were carried out at Near Beginning of Cycle (NBOC), Middleof Cycle (MOC) and End of Cycle (EOC) core conditions.
•

The fuel temperature (Doppler) coefficients are only slightly more positive in the
Am-Cm core. Hence the Doppler reactivity defect from full to zero power (i.e. thepositive reactivity insertion associated with cool-down from HFP to HZP is slightly
increased, as indicated in Table 5.6.

•

The boron coefficients are slightly less negative in the Am-Cm core, consistent
with higher thermal neutron absorption, as indicated in Tables 5.7 and 5.8 for HFPand HZP conditions. Some of the calculated values are close to or marginallyoutside the nuclear design limits. Again, the violations are so marginal thatloading pattern optimisation would expect to be effective in mitigation.

•

The control rod reactivity worths are slightly modified with the particular
distribution of Am-Cm loaded fuel assemblies. Table 5.9 compares control rod
reactivity worths at HZP conditions for the two equilibrium cores. The reactivity
worths quoted correspond to all control rods inserted except for the highest
reactivity worth control rod, which is assumed stuck out of core.

•

Table 5.10 shows the impact of Am-Cm loading on shutdown margins, which
include allowances for uncertainties. The NBOC shutdown margin is more than
800 pcm smaller in the equilibrium Am-Cm core. However, the EOC shutdown
margin is actually slightly higher in the Am-Core than it is at NBOC. This contrasts
with the equilibrium UO2 core, which shows the behaviour whereby the shutdown
margin becomes more limiting at EOC conditions. The much smaller impact of
Am-Cm at EOC is presumably because of the much lower inventory of Am
remaining at EOC.

•

Table 5.11 shows that the delayed neutron fractions are hardly changed in the AmCm equilibrium core.

•

Table 5.12 shows the beginning of cycle (BOC) boration limits at HFP, HZP and also
Cold Zero Power (CZP) core conditions. The relevant boron concentrations tend tobe higher in the equilibrium Am-Cm core, but all the nuclear design limits are
satisfied. RIL indicates that the control rods are inserted to their insertion limits.
ARO denotes all control rods withdrawn and ARI denotes all control rods inserted.
Table 5.3: HFP radial peaking factor F∆H
Cycle burn-up (GWd/t)

Eq UO2 core

Eq Am-Cm core

0

1.463

1.541

0.15

1.465

1.537

2

1.383

1.495

4

1.369

1.502

6

1.384

1.517

8

1.455

1.537

10

1.477

1.511

12

1.449

1.484

14

1.422

1.468

16

1.400

1.454

17.1

-

1.445

18.3

1.369

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5. FUEL AND CORE DESIGN

Table 5.4: HFP moderator temperature coefficient (pcm/°C)
GWd/t

EQUO2

AMCM

NBOC

-33.7

-38.1

MOC

-42.3

-48.6

EOC

-70.8

-74.4

Table 5.5: HZP moderator temperature coefficient (pcm/°C)
GWd/t

EQUO2

AMCM

NBOC

-10.8

-11.0

MOC

-11.7

-20.1

Table 5.6: Doppler defects from HFP to HZP in pcm
GWd/t

EQUO2

AMCM

0.15

985

1025

EOC

855

886

Table 5.7: HFP boron reactivity coefficients and nuclear design limits (pcm/ppm)
GWd/t

EQUO2

Limits

AMCM

Limits

NBOC-6.9-11.5,-6.3-6.4-11.5,-6.2

MOC

-7.3

-11.8,-6.6-6.6-11.8,-6.7

EOC

-8.3

-12.5,-7.5-7.4-12.5,-7.5

Table 5.8: HZP boron reactivity coefficients and nuclear design limits (pcm/ppm)

50

GWd/t

EQUO2

Limits

AMCM

Limits

NBOC-7.1-10.9,-5.6-6.6-10.8,-5.5

MOC

-7.4

-6.8

-11.2,-5.9

EOC

-8.4

-7.4

-11.8,-6.6-11.2,-5.9-11.8,-6.7

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Table 5.9: HZP control rod worths in pcm (minus the highest worth rod)
GWd/t

EQUO2

AMCM

0.15

6116

5389

EOC

6247

6545

Table 5.10: Best-estimate shut-down margins (pcm)
GWd/t

EQUO2

AMCM

0.15

4520

3625

EOC

3483

3784

Table 5.11: HZP delayed neutron fraction
GWd/t

EQUO2

AMCM

BOC

0.006237

0.006220EOC

0.005286

0.005263

Table 5.12: Boration limits
Core Condition

Design limit

EQUO2

AMCM

HFP BOC, no Xe, rods at RIL

2 000

1 361

1 469

HZP BOC, ARO, No Xe, k=0.99

2 250

1 499

1 654

CZP, BOC, No Xe, ARO, keff = 0.99

2 500

1 764

2 014

CZP, BOC, No Xe, ARI, keff = 1.0

1 600

1 075

1 396

CZP, BOC, No Xe, ARI, keff = 0.95

2 500

1 608

1 999

5.3 Material balance
Figure 5.1 shows the material balance for americium and curium for the target rods.
The blue bars show the initial weight percentage of the various isotopes in the Am-Cm
target rods, while the red bars show the weight percentages after the fuel assembly inwhich the Am-Cm rods are inserted has been irradiated to its 60 GWd/t mean discharge
burn-up. The grey bars show the difference between end-of irradiation and beginning of
irradiation. The americium isotopes all show net destruction by fission or captures (thisalso applies to 242Am, which is at too low a concentration to show on this scale. In

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5. FUEL AND CORE DESIGN

contrast, all of the curium isotopes show net in-growth. Overall destruction of americium
and curium amounts to approximately 70 w/o of the initial Am-Cm mass.
Figure 5.1: Material balance for americium and curium target rods
Isotopic evolution in Am-Cm target rods
8.00E+01

0 GWd/t
6.00E+01

60 GWd/t
DELTA

Weight percent

4.00E+01

2.00E+01

0.00E+00

Am-241

Am-242

Am-243

Cm-242

Cm-243

Cm-244

Cm-245

Cm-246-2.00E+01-4.00E+01-6.00E+01-8.00E+01

5.4 Fuel matrices
It is usually considered beneficial to use an alternative diluent matrix to UO2 forminor actinide target fuel. It is important to disperse any transuranic materials in amatrix, because the power density would otherwise be excessive. For example, becauseof its high fission cross-section, plutonium needs to be diluted by a factor of 10 or so, toobtain a manageable fission density and normally this is achieved by dispersing it in a
depleted UO2 matrix. Even though neptunium and americium do not fission as readily,
the heat production from fissions and neutron captures combined can be excessive
unless there is a matrix.
Inert matrix fuel (IMF) is beneficial to use as the matrix because the materials used
are intended to avoid neutron interactions and in particular fertile captures from 238U to239
Pu are avoided. Since the objective is to destroy minor actinides, generating 239Pu fromfertile captures is unhelpful as the freshly produced 239Pu provides the basis forgenerating new minor actinides. IMF avoids this counterproductive step. Candidate
materials include cerium oxide (or ceria CeO), zirconium oxide (or zirconia ZrO) or
yttrium stabilised zirconia (YSZ). The IMF matrix needs to be resistant to hightemperatures, corrosion and irradiation, compatible with the fissile material and fission
products and preferably have a low neutron capture cross-section. There are many
materials that are being researched and many are promising, but any selected forcommercial application will need to be proven with extensive irradiation testing.
Proving the suitability of IMF for commercial reactor application is one of its maindrawbacks. Another potential concern is that the absence of 238U resonance captures
significantly lowers the effectiveness of the Doppler feedback mechanism. However, thedeleterious effects of IMF are thought to be manageable if the core fraction of IMF fuelscontaining minor actinides is limited, as has been demonstrated in recent studies [10].

52

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Another candidate material is thorium oxide (or thoria ThO2), which is not an inert
matrix, but nevertheless is useful in this context. Natural thorium is comprised entirelyof 232Th and undergoes neutron capture to 233U, which is fissile. As a matrix material for
MA-TRU fuel, 233U may make a useful contribution towards the neutron balance,compensating for the negative neutron balance of the minor actinides. Using thoria as
the matrix may therefore help reduce the cycle length/initial enrichment penalty noted
earlier in Section 5.2.2. Despite this fertile production mechanism, thoria is stillcompatible with efficient minor actinide burning, because 233U has a very low capture
cross-section and in the majority of neutron interactions it fissions. Only a very small
proportion undergoes neutron capture to 234U and, in turn, this has a relatively low
capture cross-section. Compared with 238U as the seed for transuranic production, 233U
produces orders of magnitude less partly because the 234U cross-section is low and partly
because the capture chain starts of 5 mass units smaller. For example, to produce 241Am
from 238U requires three neutron capture events, while from 233U eight neutron capture
events are needed. With the low capture cross-sections for 233U and 234U, needed to start
the process, the probability of accumulating eight captures is vanishingly small.

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6. OPERATIONS, SAFETY AND LICENSING

6. Operations, safety and licensing

6.1 Initial fissile loading
Referring to the fission and capture cross-sections of the minor actinides in Table 1.2,
it is very clear that all the majority of the minor actinides nuclides make a negative
contribution to neutron balance. When a fission event occurs ν new neutrons are born
and the neutron that initiated the fission is lost. Therefore the contribution to neutronbalance from fission events is proportional to (ν-1)σf (σf is the fission cross-section) andthe contribution of neutron captures is -σc (σc is the capture cross-section), giving a total
contribution (ν-1)σf -σc. Although there are some compensation mechanisms, suchneutron captures on 238Pu and 244Cm to produce fissile 239Pu and 245Cm respectively, theoverall neutron balance is still negative.
A negative neutron balance implies that the minor actinides must be supplied withneutrons in a thermal reactor and the source of these neutrons is the driver fuel. In thecase of LEU driver fuel, this means that the initial 235U enrichment must be increased. If
plutonium fuel is used as the driver, a higher initial plutonium loading is required.
Failure to increase the initial fissile loading implies that in a batch loaded reactor thecycle length will be shorter and the discharge burn-up reduced.
With LEU fuel, increasing the initial fissile loading is a cost penalty because of theadditional uranium ore, conversion and enrichment requirements. Alternatively, a utility
might choose to accept a shorter cycle length, but this involves an operational cost
penalty, as will be discussed in Chapter 7.
With plutonium fuel, increasing the initial fissile content incurs no cost penalty, but
the total plutonium content is usually limited by the need to maintain a negative void
coefficient. In conventional MOX fuel the void coefficient normally becomes positive
when the total plutonium content is ~12 w/o. Increasing the initial plutonium content tocompensate for the negative neutron balance of minor actinides risks encroaching on thevoid coefficient limit.
The study discussed in Section 5.2 illustrates just how much of an impact thenegative neutron balance of minor actinide core loading has on a PWR core. Although theresults are specific to the particular reactor and core loading pattern, they still provide a
useful reference point because the spectral characteristics of different PWRs (and evenBWRs) are not radically different and as an indicator of the order of magnitude effect, theresults are transferable.
Recall from Section 5.2 that the equilibrium 18 month PWR fuel cycle was loaded withslightly over 100 kgHM of Am-Cm target rods. The initial enrichment of the UO2 driver fuelwas kept constant and the 3-D nodal whole core model indicated a reduction in equilibrium
cycle length from 18.3 GWd/t to 17.1 GWd/t. It is estimated that the 18.3 GWd/t cycle length
could be restored by increasing the mean initial 235U enrichment of the entire reload
batch by 0.26 w/o. This is a modest penalty that could easily be accommodated
technically (provided the 5.0 w/o fabrication limit is not exceeded), though at a cost to theutility. Chapter 7 shows how this translates into an economic penalty.

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6. OPERATIONS, SAFETY AND LICENSING

6.2 Impact on plant operations
This section reviews the impact of MA-TRU fuels on operations for an LWR utility.
Much of the discussion, however, will also be valid for other thermal reactor types suchas CANDU and HTR where minor actinide loading might be adopted.

6.2.1 Fresh fuel receipt
The first area affected is fresh fuel receipt and handling. Conventional LEU or MOX
fuel is normally delivered to the reactor site in unshielded transport containers and the
fresh fuel can be handled safely while it is transferred to the fuel ponds. No radiologicalshielding measures are required for LEU or MOX fuel, though there may be measurestaken to minimise exposure to the gamma emissions from MOX fuel. The fresh fuel may
be subject to inspection as part of the process before it is loaded in the fuel pond.
MA-TRU fuels would be transported in shielded transport containers and the utility
would need to put arrangements in place so that it is able to receive shielded containers
and open them safely. It may be necessary to build a new container receipt area and thiswould need to be equipped with a crane that can handle the additional weight. The
utility would also need to install any physical shielding that may be necessary and put
into place radiological operating procedures to protect workers during fresh fuel receipt
and transfer operations. The modified fuel receipt facility and the operational procedures
will need to be approved by the licensing authorities, taking account of any limitations
due to heat load during transfer to wet storage, as noted earlier in Section 3.5.
The utility will incur an initial investment cost on account of the changes needed andpossibly may incur an additional operational cost every time a delivery of MA-TRU fuels
is made. Another potential cost element for the utility would be if it is necessary to
upgrade the security arrangements at the reactor site. The security requirements neededfor MOX fuel are already more stringent than those for LEU, because of the increased risk
posed by plutonium-bearing fuel. With MA-TRU fuels the source term following aterrorist attack is likely to increase further and a review of security arrangements would
be necessary to ensure the appropriate level of protection is implemented. This
represents another potential operational cost increase to the utility. Bearing in mind that
LWRs are refuelled off-line, it is possible that more stringent security arrangements
would only need to be in place for the period when MA-TRU fuels are first received andstored in the fuel ponds. Once the MA-TRU fuel is transferred to the core and irradiated,
it poses no additional threat.

6.2.2 Core loading
Core loading pattern optimisation will be performed prior to each operating cycle. As
noted in Section 5.2, the studies carried to date indicate that MA-TRU fuel assemblies canbe accommodated mostly within the normal operating envelopes and acceptable loading
patterns are therefore considered feasible.
Other issues that a utility would need to address prior to irradiating MA-TRU fuelassemblies would include:56•

review of the core loading strategies that ensure sub-criticality during thereloading procedure;•

review and analysis of fresh fuel handling accident scenarios, including
misloading events;•

review of procedures for approaching criticality, including neutron detector source
range settings. It may be necessary to determine whether MA-TRU fuels with highintrinsic neutron sources will have any impact on source range detector
sensitivities.

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6. OPERATIONS, SAFETY AND LICENSING

In addressing these areas, the utility would incur an initial investment and licensing
cost, followed by (presumed) smaller on-going operational costs during each refuelling
operation.

6.2.3 Core operation
During the actual operation of the core, the core characteristics would be affected by
the presence of the MA-TRU fuel, but if the nuclear design limits are satisfied there is no
reason to suppose the core would behave in any untoward manner. There would,
however, be an increased risk of the core-follow data not conforming with expectation,
especially for the first cores to be loaded with minor actinides. There is no operational
experience of LWRs burning MA-TRU fuels and nuclear data libraries cannot be relied
upon to the same extent with MA-TRU fuels as they can for conventional fuels. These
uncertainties might manifest themselves in the core departing from the expectedbehaviour and this is a risk that the utility would be exposed to. This would be addressedwith a strategy of initially loading small numbers of minor actinide rods in a core,followed by a progressive increase of minor actinide loading with an associatedmonitoring programme to ensure acceptable behaviour and provide code validation data.
With any new fuel design, there is always an elevated risk of unexpected fuelbehaviour, perhaps leading to premature fuel failures. This will be an issue for MA-TRU
fuels, especially so for the first commercial-scale loadings. Again, it is the utility that will
be exposed to the risk of fuel failures and the costs associated with investigating fuelfailures and possibly premature discharge of the fuel assemblies containing failed rods.
Conventional LEU and MOX fuels benefit from a very expansive operational experiencebase that will initially be lacking for MA-TRU fuels. Even as experience accumulates, it islikely that MA-TRU fuel irradiation experience will always lag behind LEU and MOX, sothat there will remain an elevated risk of fuel failures.

6.3 Code validation
The core design analysis to date has been carried out using the standard nuclear
design codes validated for LEU and MOX cores. In LEU fuel the inventories of the minor
actinides are only of the order of 0.1% by mass and any nuclear data uncertainties for theminor actinides would only be expected to have a minor impact on in-core performance.
For MOX fuels the minor actinide inventories are considerably higher, but so far there has
been no indication that minor actinide cross-sections are affecting the predictive powerof the core design codes. However, it is always possible that unknown cancellation effects
may be occurring that are masking any biases in minor actinide data and that theunderlying biases may become evident at higher loadings.
Minor actinide nuclear data has already been identified as a priority area for nuclear
data R&D [11]. New cross-section evaluations for minor actinides will need to be followed
up by sensitivity analyses to determine the effect on integral core parameters. The new
evaluations will eventually need to be tested against the first commercial-scale minor
actinide irradiations to establish the validation base for the nuclear data libraries andnuclear design codes. This can be expected to be a prolonged activity before the nuclear
design methods can be considered mature for minor actinide applications.
Unexpected nuclear design behaviour is another risk that will fall on the utilities. At
worst, the utility would risk losing output if there was, for example, a need to temporarily
down rate or possibly suffer a shortfall cycle length due to insufficient reactivity.

6.4 Licensing timescales
The industrial utilisation of new types of fuel requires detailed preparation from thetechnical as well as from the licensing viewpoint. In particular, the licensing situation of all

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6. OPERATIONS, SAFETY AND LICENSING

concerned areas has to be reviewed: transportation, manufacturing, design, and irradiation in
the reactor. Acquiring new licenses or the adaptation of existing licenses may be a tediousand time consuming process and should start early in the project.
A general sequence of steps for licensing of new types of fuel after specification of the
manufacturing process is described in Table 6.1.
Dependent on the manufacturing, design and irradiation experience with similartypes of fuel Phase I and II may be overlapping. A reduction of the time needed forindustrial implementation of MA-TRU fuel is dependent on the fuel design. If the
admixture of minor actinides to the new fuel expected to have only minor impact on thethermal-mechanical behaviour, then the required tests can partly be performed inparallel. In any case, the time for commercial implementation of new types of fuel isconsiderable and dependent on expert opinion on the necessity for test irradiations it isin the range of minimum 10-15 years.
Table 6.1: Licensing of MA/TRU fuelCharacterisation oflicensing phases

Phase I: (1-2 years;>5 years if
irradiation tests are
required)Phase II: (> 7 years
includingmanufacturing, post
irradiation
examinations)Definition of boundary
conditions forintended use of
MAThR fuel (type ofreactor, operationmode, etc.).
Evaluation of licensing
requirements

Manufacturing

Determination of
manufacturing facility
and check for required
adaptations ofspecifications,guidelines, extensionsof licenses and of
manufacturing
equipment;
Manufacturing of testpellets
Adaptation of
manufacturing
equipment,
specifications andguidelines;

Preparation oflicensing

Initiation of thelicensing process for
manufacturing;
Qualification of the
manufacturing
process;
Manufacturing of fuelpellets and rods in
LTA quantities

Phase III:
(1-2 years)

Licensing

Phase IV:

Commercial
implementation

58

DesignIrradiation

If deemed necessary:
Irradiation of test fuelrods for qualificationof thermalmechanical andnuclear design
parameters;
continuous
surveillance and postirradiation
examinations

Determination of- design parameters(e.g. fuel thermal
conductivity, radial
power profile) from▪ literature▪ manufacturing tests▪ test irradiations- applicability ofdesign codes

Identification of
possible shipping
casks.

Qualification ofthermal-mechanicaland nuclear design
parameters (e.g. fueldimensional stability,
thermal conductivity,
power profile, nuclear
cross-sections) anddesign codes;

Update of- licensing status of
the shipping/storagecasks forrequirements arisingfrom the new fuel- existing guidelines
for shipping andhandling(manufacturing site,
plant site, storagesite)

Insertion of Lead Test
Assemblies (LTAs)
for genericdemonstration ofirradiation behaviour;
continuous
surveillance and postirradiation
examinations

If required: licensingof the shipping cask

Licensing of insertionof fuel assemblies inreload batch size

Adaptation of
specifications

Thermal mechanicaland neutronic designof pellet, fuel rod andfuel assembly
Manufacturing of
MAThR fuelrods/assemblies

Transportation,
handling,
reprocessing, longterm storage

Update of genericdesignIdentification ofboundary conditions
for reprocessing
and/or long-term
storage

Insertion of final
design MAThR fuelassemblies in batch
size

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7. ECONOMICS

7. Economics

7.1 Costs and benefits
The economics of minor actinide recycling in thermal reactors is a complex subject
for which there are, at present, no definitive answers. Nevertheless, it is important toconsider the economic issues here, even if only qualitatively, to highlight what questions
will need to be addressed before minor actinide recycling can be implemented in thermal
reactors.
When introducing a new fuel design or other fuel cycle innovation, a utility would
normally expect to bear an additional cost in implementing the change, but would expect
to demonstrate a tangible benefit in return. For example, an increase in discharge burnup would incur extra initial cost for the fuel because of the higher initial enrichment
required. In return, the utility would reduce its fuel throughput and possibly make otheroperational savings that would compensate.
There are clear direct costs to the utility associated with minor actinide recycling thatwill similarly need to be balanced against the benefits. However, a complicating factor forminor actinide recycling is that the direct costs will need to be met by the utility, while
the benefits may not necessarily accrue to the utility, but rather to society as a whole.
Another complicating factor is that the benefits may be in the form of reduced futureradiological dose commitments and a mechanism is needed to equate the dose reduction
to an economic saving. This raises the possibility that mechanisms may need to be foundwhereby the utility is compensated for its direct costs through subsidisation mechanisms
from governments.
The additional costs that a utility would incur if it decided to recycle minor actinides
in a thermal reactor would be:
1. The cost of separating the minor actinides during reprocessing and incorporatingthem in the nuclear fuel, in either homogeneous or heterogeneous MA-TRU fuels.
2. The additional costs associated with transport of MA-TRU fuels to the reactor site.
3. The additional costs associated with the safe handling and transfer of the MA-TRU
fuel to pond storage.
4. Any direct cost of the additional fissile material needed to compensate for theminor actinide loading. For a uranium fuelled thermal reactor, a higher initialenrichment would be incurred, which implies a higher cost for uranium
procurement and enrichment services. For a plutonium-fuelled thermal reactor, it
is possible that a compensatory increase in initial plutonium loading will be
required and that no additional direct cost will be incurred if the plutonium isowned by the utility.
5. The additional costs associated with licensing the reactor for operation with MATRU fuels.
6. Any additional costs associated with discharge, storage and transport of MA-TRU
fuels.
7. Any additional costs associated with management of spent MA-TRU fuels, whichmight be stored and conditioned for disposal or recycled further. This would
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7. ECONOMICS

include the management and disposal of any additional waste streams generated
by management of spent MA-TRU fuels.
Balancing these costs are the benefits of minor actinide recycling:(a)

The avoided future cost of repository construction and operation that can beviewed principally as an economic benefit (through there may be other beneficial
aspects, such as increased political acceptance). This might be evident as a direct
economic benefit to the utility if its spent fuel costs were reduced. However, witha levelised cost methodology, the spent fuel costs will incur a discounting effectover many years such that such a reduction may not have a noticeable effect on
the levelised fuel cycle cost.

(b)

Minor actinide recycling will involve an increase in radiological doses to workers
during separation, fuel fabrication and all the subsequent operations involving
MA-TRU fuels. On the other hand, recycling of minor actinides may reduce thefuture radiological dose commitment from geological repositories. If it can bedemonstrated that the future avoided dose outweighs the additional operationaldose commitment, then this would constitute a benefit. Any avoided futureradiological dose uptake represents a societal benefit that could be regarded as
an economic gain.

It is important here to try to estimate at least the order of magnitude of the costs
associated with items 1) to 7) above would be and to balance these against the economic
benefits (a) and (b). This is an important input to the debate that will need to take place
between utilities and government before minor actinide recycling can be realised. It
should be understood that minor actinide recycling is currently very far from commercial
realisation and that any economic analysis will inevitably be very imprecise at this stage.
Definitive economic estimates will not be available until minor actinide recycling has
actually been implemented at commercial scale, at which point construction andoperating costs for the associated fuel cycle plants will be better known. Nevertheless,
Section 7.2 attempts to provide some indicative estimates of the order of magnitude costs
that the utility will incur.

7.2 Cost impact
Each of the seven areas where minor actinide recycling would incur additional costs
is considered here in turn. In some cases it is premature to attempt to make aquantitative estimate and all that can be done here is to make some qualitative
observations. The overall cost analysis will therefore be incomplete, but the overall
analysis is still worthwhile because it quantifies at least some of the costs and identifies
where the gaps are:7.2.1 Separation
It is not possible to estimate the cost of a minor actinide separation plant at this time.
No commercial scale plant designs have been developed for a reprocessing plant with aminor actinide separation capability and there are no engineering estimates of the capitaland operating and maintenance (O&M) costs. Moreover, it might be difficult in practice todisentangle the cost of the minor actinide separation facilities from the overall costs of
the reprocessing plant. A more valid approach might be to consider the entire plant as asingle entity and concentrate instead on demonstrating that the overall back-end costs
are competitive with conventional reprocessing plants. This suggests that the problem
should perhaps be inverted and the need for economic competitiveness to be imposed asa requirement on the design of the integrated plant. While it may be acceptable for a
reprocessing plant capable of separating minor actinides to have a higher unit cost than aconventional reprocessing plant, any such cost penalty should be kept within reasonablebounds. Reprocessing unit costs indicated in the 1994 OECD/NEA study of back-end costs

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are of order of 1 000 USD per kgHM. It would perhaps be sensible to aim for any cost
penalty associated with minor actinide recycling to be no more than a few hundred
USD/kgHM, or justifying it to utilities will be very difficult.

7.2.2 Transport
No transport packages have been licensed for MA-TRU fuels and it is not possible atthis time to estimate the effect on transport costs. It is possible that existing spent fueltransport packages might be adaptable for carrying MA-TRU fuels, which would avoid theneed to develop and license new package designs. It would be very beneficial if the same
packages used for transporting spent fuel assemblies to the reprocessing plant could also
be used to ship MA-TRU fuel assemblies on the return trip, rather than being shipped
empty. Whether this is possible and what loadings of MA-TRU fuels could be permitted
depends on the specific details of the transport package designs and in particular theneutron source strength of the MA-TRU fuels. It will have to be sufficient for now to notethat spent fuel transport, although expensive in absolute terms, only represents a verysmall fraction of the fuel cycle cost and even if a substantial multiplier was necessary for
MA-TRU fuels, it is unlikely to have more than a minor overall impact.

7.2.3 Fresh fuel receipt and storage
The utility would need to install any additional equipment needed for receipt,
inspection, handling and storage of un-irradiated MA-TRU fuels and to license suchoperations with the licensing authority. For heterogeneous minor actinide target fuels,
the utility would need to provide equipment and personnel for receipt and storage,
loading and un-loading the targets in-core during refuelling operations. The initial
equipment installation is likely to represent a substantial cost perhaps a few million
dollars and it seems reasonable to allow for an order of magnitude allocation of USD 12 m per reload for heterogeneous targets and perhaps less for homogeneous MA-TRU
fuels.

7.2.4 Core reactivity effect
The presence of MA-TRU fuels in the core will have an impact on the nuclear design
behaviour of the core. A recent study of americium-curium targets loaded in a PWR core[10] suggests that worthwhile loading of 100 kg of minor actinides could beaccommodated in a PWR while satisfying the normal nuclear design limits. This will
ensure that the nuclear design performance of the core will not be a factor that will limit
core output. However, the analysis in [10] does confirm that the presence of minor
actinides in the core depresses core reactivity, requiring a compensatory increase ininitial fissile loading. For a uranium fuelled core, this implies increased uranium ore andenrichment procurement costs which can be evaluated with reasonable confidence.
The utility could compensate for the reduction in core reactivity by increasing theinitial enrichment of the fuel. This appears as a direct cost penalty through the increasein uranium ore conversion and enrichment costs. In terms of minimising the operational
penalty, this is the preferred option, because it preserves the nominal cycle length andthe utility does not lose out on generating revenue. The enrichment penalty equates toan increase in the initial 235U enrichment of 0.26 w/o averaged across the entire 84 reload
batch, as indicated in [10]. Assuming a nominal mean enrichment of 4.2 w/o, a uraniumore price of 130 USD/kg, conversion at 10 USD/kg, enrichment at 90 USD/kgSWU, a 5%
discount rate and 0.5% losses this equates to a direct cost penalty to the utility of
USD 5.6 m per 18 month reload. This would be very significant to a utility.

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7.2.5 Licensing
The utility would need to submit a safety case to the licensing authority todemonstrate prior to loading MA-TRU fuels. This would need to cover all aspects ofoperations involving MA-TRU fuels from receipt and storage, core loading irradiation,discharge, storage and transport. The safety case would need to show that the presenceof MA-TRU fuels would not impact safe behaviour of the fuel and core in normal
operation and in fault conditions. The cost of generating the initial safety case is likely to
be significant, perhaps several million US dollars. Spread over 10 years and allowing forother costs associated with licensing such as fuel monitoring programmes todemonstrate satisfactory performance of the MA-TRU fuels, it would be reasonable toassign an order of magnitude allocation of USD 1-2m per reload.

7.2.6 Discharge, storage and transport of spent MA-TRU fuels
There is the possibility that spent fuel handling operations at the reactor may be
impacted by MA-TRU fuels. Spent fuel is discharged into the fuel cooling ponds, where it
will remain until such time as it is transferred to dry storage or transported to a separate
spent fuel facility for interim storage/disposal or reprocessing. There is the possibility
that spent fuel management cost at the reactor site may be impacted by the different
characteristics of MA-TRU fuels. In particular, transport costs may be impacted if the MATRU fuels have a higher heat output or neutron output than conventional fuel assemblies.
The possible cost implications cannot be estimated at present, because the necessary
R&D work has not been carried out.

7.2.7 Spent fuel management
Spent fuel management costs may be affected by MA-TRU fuels. Spent MA-TRU fuelthat is handled along with conventional fuels in a fuel conditioning plant or reprocessing
plant may have a cost impact. It may be necessary to make plant modifications toequipment and possibly to operational procedures, both of which could affect unit costs.
It is possible that plant throughputs would be limited while handling MA-TRU fuels,
which could have a major impact on unit costs. Alternatively, where MA-TRU fuels
needed to be handled in separate facility, the entire investment and operating cost ofsuch a facility would need to be supported either by the utilities using it or by thegovernment. Since no designs for spent fuel conditioning or reprocessing plants capableof handling minor actinides have been developed, it is premature to attempt to make aquantitative estimate at this time.

7.2.8 Overall cost impact
These are just very crude and incomplete estimates, but they amount to aquantifiable economic cost to the utility of the order of USD 10 m per reload, which issubstantial. This estimate excludes several other cost components which are also likely
to be substantial and so should be regarded as a lower bound figure that is intended only
to inform the debate at this early stage of development.

7.3 Economic benefits
7.3.1 Repository cost savings
It is possible that as a direct result of minor actinide recycling, the future capacity of
the geological repository may be reduced or that capacity may be deferred for a numberof years. In such a scenario, there would be an economic benefit to the organisation;
usually the government that is responsible for funding the construction and operation of
the repository. It is not possible to attempt a meaningful estimate of how much of asaving this might represent, not least because any such saving is likely to be very
dependent on the specific scenario that might apply to a given country. In theory, a62

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utility might benefit by seeing a reduction in its spent fuel disposal levy, but given thelong lead times involved, discounting effects are likely to make the benefit very marginal.
It is possible that the socio-political benefits might actually outweigh any direct
economic effect. It is conceivable that minor actinide recycling might be viewed as
removing a potential barrier to sustainability if it reduces the repository capacity needed
or defers capacity until later.

7.3.2 Avoidance of radiological doses
The economic equivalent of avoided future radiological doses can in principle becalculated given knowledge of:•

The reduction in radiotoxicity results from minor actinide recycling. Once the fuelcycle scenario has been defined, this can easily be evaluated using fuel cycle
scenario analysis codes ORION [12] and NFCSS [13].

•The transfer coefficients of radionuclides from the geological repository tobiosphere as a function of time. This is much more difficult to evaluate, as it
depends on the specific geological and hydrological characteristics of a particularrepository.

Calculating the economic benefit of avoided future radiological doses will require abetter understanding of the transfer coefficients and also requires agreement as to how
to assign an economic value to avoided dose i.e. what is the equivalent economic cost of
1 man-Sievert of exposure. There have been various attempts to estimate the Svequivalent cost in the literature and different valuations have been arrived at. Reference[14] makes an important distinction between 1 man-Sievert of exposure to radiation
workers, who choose to accept the small risk associated with occupational exposure andmembers of the public exposed to aerial and water-borne exposure without being able tochoose otherwise. The work reported in [14] arrived at a weighting factor to apply to
public exposure of between 2 and 6 times that which should apply to radiation workers.
This was based on an attitude survey from an opinion poll and although the precise
values to use are debatable, the principle appears sound. The weighting in favour of thecost of public exposure will benefit future avoided dose uptake over operational dose
commitment.
At the present time, the lack of understanding of radionuclide migration from
repositories prevents quantitative estimates of the economic benefit from avoided future
dose. Any such assessment is likely to be different for each specific country’s scenarios
and requires a consensus to be established as the economic equivalent value of avoided
dose.

7.4 Subsidisation mechanisms
Section 7.2 has quantified some direct costs that a thermal reactor operator would
incur in recycling minor actinides in a thermal reactor. Although incomplete, those costs
which have been quantified (however approximately) are quite significant. On the other
hand, Section 7.3 has identified potential benefits, which though not quantified, may
potentially be sufficient to counter balance the costs. However, the important point tobear in mind, is that while the utility will incur additional costs, the benefits are likely to
be in the form of indirect societal returns. In these circumstances, it may be necessary forgovernments to put in place subsidisation mechanisms to encourage thermal reactor
operators to recycling minor actinides.
For a private company, agreement of the level of subsidisation would be a matter forcommercial negotiation. At the very least, a utility might expect its immediate direct
costs to be covered and other factors might also influence the negotiations. For example,

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there would be an element of risk associated with minor actinide recycling and a utility
might demand compensatory payment to cover it.
The overall conclusion is that subsidisation mechanisms are likely to be necessary to
encourage thermal reactor operators to recycle minor actinides. At present, it is notpossible to fully quantify the competing effects and further R&D is required to quantify
on the one hand the additional costs and on the other hand, the economic equivalentbenefits.

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8. Research and development needs

This chapter discusses the research and development requirements for minor
actinide recycling in thermal reactors. It follows the flow of minor actinides through eachfuel cycle stage, starting with separation of the minor actinides from spent nuclear fuels
during reprocessing:

8.1 Separations
Flowsheet development will be needed for the selected minor actinide separations
technology. R&D is needed to take separations technology from its present immaturestate to commercial readiness. Facilities work would initially be carried out at laboratory
scale and this would be followed by a sub-commercial scale development/demonstration
facility, leading eventually to the construction and operation of the full scale plant. The
minimum timescale for commercial implementation is of the order of 20 years.
It would be useful if at an early stage conceptual separation plant designs could be
developed which would be useful for informing the justification process and provide
initial indications of capabilities, limitations. Of particular interest would be tocharacterise the waste streams from the minor actinide separation process and check
whether they would be compatible with existing waste streams and disposal
requirements.
Most R&D has been focused on the initial separation of minor actinides from spent
fuel. However, another important aspect will be the subsequent management of
irradiated minor actinide target fuels, which may involve conditioning for disposal or
further separation and recycling steps. It is important that R&D on target management is
not neglected.

8.2 Fuel fabrication
R&D is needed to develop the selected fuel fabrication technology to commercial
readiness. A small number of laboratory scale fuel fabrication facilities are alreadyavailable that could carry out small scale development work. Further progress would
require a demonstration scale fuel fabrication facility, leading to the design andconstruction of a commercial scale plant. The timescale to commercial readiness isestimated at 10 to 15 years, as noted earlier in Section 6.4.
As, with separations technology, it would be useful for justification if conceptual fuelfabrication plant designs were available at an early stage.

8.3 Fuel transport
The technical requirements for transport of un-irradiated and irradiated MA-TRU
fuels will need to be developed. It would be very helpful if existing transport container
designs could be adapted for transport of MA-TRU fuels. An early evaluation of thefeasibility of using existing transport container designs for MA-TRU fuels would be
helpful to establish any limitations and provide initial indications of costs. Theoreticalstudies could be completed on a short timescale, perhaps 2 to 3 years and provide an

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useful indication of limitations and potential costs. Development and licensing of new or
modified transport containers would require perhaps 5 to 10 years, but is unlikely toproceed until there is a clear commercial requirement.

8.4 Fuel design
R&D is required to progress MA-TRU fuel designs from the conceptual stage, where
they are at present, to commercial readiness. Further research is needed into materials
properties for MA-TRU fuels to provide the information needed for the fuel performancecodes which are used to demonstrate satisfactory thermo-mechanical fuel behaviour.
The fuel behaviour codes will require validation against experimental measurements andfrom post irradiation examination (PIE).
Experimental work will be required in the following areas:•

Fuel materials properties measurements.

•

Small scale irradiation tests of MA-TRU fuels in test reactors, which will require
suitable fuel fabrication facilities to be available and qualified to handle minor
actinides in small quantities (up to a few kg).

•

Small scale irradiation trials in power reactors (small numbers of individual minor
actinide rods) leading to intermediate scale irradiation trials (small numbers ofminor actinide assemblies) and eventually to full scale commercial loadings.

The timescale for irradiation testing of MA-TRU fuels and validation of fuelperformance codes is of the order of 10 to 15 years, with a further lead time of perhaps
5 years for test fuel fabrication, giving an overall programme of 15 to 20 years.

8.5 Core designCore design R&D is required to develop viable core loading patterns and fuelmanagement schemes for MA-TRU fuels and to validate nuclear design methods. Minor
actinide nuclear cross-section data will need to be improved and this may require newexperimental measurements and nuclear data evaluation effort. Validation of nuclear
design methods may necessitate integral measurements in zero power facilities, followedby small-scale irradiation trials in commercial reactors. The overall programme would be
similar to that for fuel design, 5 years fabrication followed by 10 to 15 years to progress tocommercial scale loading.

8.6 Spent fuel characterisation
Spent fuel characterisation data will need to be generated for MA-TRU fuels. This
includes decay heat and neutron source data, which is important for spent fuel storage
and transport, reprocessing and disposal. The inventory codes used to generate decay
heat and neutron source data will need to have their validation base extended to cover
the higher minor actinide inventory. The improved cross-section data needed for core
design will also be applicable to the inventory codes, but improved nuclear data decay
data may also be required. PIE of MA-TRU fuels irradiated in test reactors may be needed
to validate the inventory codes. The timescale matches that of core design R&D, 15 to
20 years to progress to commercial core loadings.

8.7 Fuel cycle assessment
Fuel cycle R&D is needed to provide a full understanding of minor actinide recycling
to inform the justification process using the full range of assessment tools that arebecoming increasingly prevalent worldwide. Useful preparatory work could begin

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8. RESEARCH AND DEVELOPMENT NEEDS

immediately, but improved assessments will be possible once the other R&D work above
starts to produce data which will better characterise minor actinide plants. The
justification of minor actinide recycling is not as straightforward as justifying new build
plants, whose main benefit is the value generated by electricity production. It will beessential to be able to support clear justification statements for minor actinide recycling,which requires a very good understanding of the benefits and detriments and the balance
between them. Such assessment work will be required throughout the period leading up
to the commercial application of minor actinide recycling.
The fuel cycle assessment will need to encompass technical aspects to do withmaterials flows, radiological dose uptake to workers and the public, long term
radiotoxicity, repository heat load and economics. Economics assessment is a major partof the justification and will need to be developed as the conceptual plant designs evolveand the available estimates of plant construction and operational costs improve.

8.8 Overall timescales
From the discussion above, it is clear that the minimum timescale for development ofminor actinide recycling in thermal reactors is of the order of 15 to 20 years tocommercial readiness. The required R&D will involve several parallel R&D streams thatwill be closely linked to one another. It is only likely to happen if an early consensus canbe established as to the benefits of minor actinide recycling in thermal reactors and aclear need agreed on. The R&D programme will need to be well funded to achieve thegoal in the 15 to 20 years and any lack of commitment will extend the R&D period for
longer.

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9. Summary and recommendations

The purpose of this study was to review the literature to determine whether thermal
reactors are a viable alternative to fast reactors as minor actinide burners. The clear
outcome of the review is that thermal reactors can indeed burn minor actinides
effectively and could play a valuable role in reducing the radiotoxicity of nuclear waste.
Although thermal reactors may in some respects be more limited in what they canachieve, they may nevertheless be valuable in this role, either as an intermediate step
towards a fleet of minor actinide burning fast reactors or as an end point in their own
right.
As an intermediate step, minor actinide burning in thermal reactors could be used todemonstrate the technologies required, and possibly expedite the deployment of minor
actinide burning fast reactors. These include minor actinide separation, MA-TRU fuelfabrication and transport and spent fuel management. Large numbers of thermal reactors
are operational today and only a few would be needed to demonstrate the technology. In
principle, there is no reason why an R&D programme on minor actinide burning inthermal reactors could not begin immediately, perhaps demonstrating the technology
within a period of 10 to 15 years. Even the most optimistic forecasts for fast reactor
deployment envisage only a few demonstration reactors being operational within such atimescale. This is one possible role of minor actinide burning in thermal reactors. A
successful outcome would demonstrate the required technology on a commercial scaleand help inform the debate as to whether minor actinide burning is necessary over thelong term.
The objective of minor actinide burning being adopted on a commercial scale incurrent and new build thermal reactors within a much shorter timescale is an end-point
in its own right. The reactors are already in operation or being built and large-scale minor
actinide burning sufficient to stabilise the minor actinide inventory could possibly beachieved within 20 to 30 years, much sooner than could be reasonably expected in fast
reactors. The capacity to demonstrate stabilisation of long-lived radionuclides would be
helpful in justifying new nuclear build, especially if many countries were to adopt
aggressive new build construction programmes.
Minor actinide burning R&D is currently still at the conceptual stage and has yet toprogress beyond theoretical studies and small-scale laboratory sample sizes. To progress
to a commercial-scale demonstration and ultimately commercial deployment will requireconsiderable commitment. The necessary steps would be:•

to develop separation flowsheets at laboratory scale and then progress tocommercial-scale separation facilities, demonstrating that these can meet therequisite radiological and environmental standards;•

to develop MA-TRU fuel designs, demonstrating these in small scale irradiation
programmes and then in commercial scale irradiation programmes. Utilities willneed to be convinced that MA-TRU fuels can be reliably irradiated;•to set up a fuel performance and nuclear design code validation programme;•

to develop MA-TRU fuel fabrication plant designs and demonstrate their viability
at commercial scale, again meeting the requisite radiological and environmental
standards.

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9. SUMMARY AND RECOMMENDATIONS•

to develop strategies for managing irradiated MA-TRU fuels safely, either in
preparation for eventual direct disposal or recycling, again meeting the requisite
radiological and environmental standards;•

to demonstrate that the radionuclide waste streams from minor actinide
separation, fuel fabrication and recycling plants can be integrated with existing
radioactive waste;•

to ensure that LWRs can achieve relative efficiency in minor actinide burning if
limited in the number of recycling and hence in scope. Transmuting in thesereactors could hence prepare a more effective minor actinide burning in fast
reactors.

The principal obstacle to the implementation of these programmes will be securing
the necessary funding from governments, since it is unlikely to come from utilities
because the benefits, as discussed in the report, will be mostly societal and not of direct
value to the utilities. Before such large scale funding can be justified, clear
demonstrations of the benefits will be required. This requires an international effort todevelop improved models of future world scenarios for nuclear growth and in particular
to indicate how the spent fuel and/or high-level waste arisings will be managed. If it canbe demonstrated that current strategies for managing high-level waste from nuclear
plants will ultimately limit the deployment of nuclear reactors, then the result will be astrong justification for minor actinide burning.

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REFERENCES

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ICRP (1995), Age-Dependent Doses to Members of the Public from Intake of Radionuclides,
Part 5 Compilation of Ingestion and Inhalation Coefficients, ICRP Publication 72, Ann. ICRP
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[2]

NEA (2011), Potential Benefits and Impacts of Advanced Fuel Cycles with Actinide
Partitioning and Transmutation, OECD, Paris.

[3]

Ronald J. Ellis (2009), “Reactor physics assessment of the inclusion of unseparated
neptunium in MOX reactor fuel”, Advances in Nuclear Fuel Management IV (ANFM 2009),
Hilton Head Island, South Carolina, USA, 12-15 April 2009, American Nuclear Society,
LaGrange Park, IL.

[4]

B Hyland, G.R. Dyck (2007), “Actinide burning in CANDU reactors”, Global 2007
Conference on Advanced Nuclear Fuels Cycles and Systems, Boise, Idaho.

[5]

B. Hyland, R.J. Ellis, G.R. Dyck, G.I. Maldonado, J.C. Gehin, G.W.R. Edwards (2009),“Transmutation of americium in light and heavy water reactors”, Proceedings of
Global 2009, Paris.

[6]

J.C. Kuijper et al. (2004), “HTR-N reactor physics and fuel cycle studies”, 2ndInternational Topical Meeting on HTR technology, Beijing, China, held on 22-24
September 2004.

[7]

J.C. Kuijper et al. (2010), “Plutonium and minor actinide management in thermal
high-temperature gas-cooled reactors”, Publishable Final Activity Report, European
Commission, PUMA Project document PUMA-1006-D411, NRG Document
21944/10.104869, Nov 2010, www.puma-project.eu.

[8]

IAEA (2009), Status of Minor Actinide Fuel Development, IAEA Nuclear Energy Series, No. NF-T4.6, Vienna.

[9]

T.A. Taiwo, T.K. Kim, J.A. Stillman, R.N. Hill, M. Salvatores, P.J. Finck (2006),“Assessment of a heterogeneous PWR assembly for plutonium and minor actinide
recycle”, Nuclear Technology, Vol 155, No. 1.

[10] G.M. Thomas, K.W. Hesketh, S. Arm (2010), “The potential of pressurised water reactors
for recycle of americium-curium”, Physor, Pittsburgh, Pennsylvania, US, 9-14 May 2010.
[11] NEA Nuclear Data High Priority Request List, www.oecd-nea.org/dbdata/hprl/.
[12] A. Worrall, R. Gregg (2007), “Scenario analyses of future UK fuel cycle options”,
Journal of Nuclear Science and Technology, Vol. 44, p.249.
[13] IAEA Nuclear Fuel Cycle Simulation System (NFCSS), www-nfcis.iaea.org/.
[14] C. Schieber, T. Schneider, L. Eeckhoudt, P. Godfroid (2000), “The monetary value ofthe man-sievert: A proposal for public exposure”, Bulletin of the Nuclear Protection
Evaluation Centre.
[15] G. Youinou, A. Zaetta, A. Vasile, M. Delpech, M. Rohart, JL. Guillet (2001),“Heterogeneous assembly for plutonium multi-recycling in PWRs: The CORAIL
concept”, GLOBAL 2001, Paris, France.

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REFERENCES[16] H. Golfier, J. Bergeron, A. Puill, M. Rohart (2000), “Multi-recycling of plutonium andincineration of americium, curium and technetium in PWRs”, Proc. Int. Conf.
Safewaste’2000, Montpellier, France.
[17] J. Bergeron, A. Puill, S. Aniel, M. Rohart, G. Rouvière (2000), “Plutonium multirecycling in PWR: The CORAIL concept”, ICONE 8, USA.
[18] G. Youinou, F. Varaine, A. Vasile (2003), “Plutonium and americium multirecycling in
the European pressurized reactor (EPR) using slightly over-moderated U-235
enriched MOX fuel assemblies”, presented at the International Conference on
Advanced Nuclear Energy and Fuel Cycle Systems (GLOBAL 2003), New Orleans, US.
[19] H. Golfier, M. Rohart, S. Aniel, J. Bergeron, J-P. Deffain (2001), “Plutonium and minor
actinides recycling in PWRs with new APA concepts”, GLOBAL 2001, Paris, France.
[20] F. Varaine, C. Chabert, G. Youinou, D. Warin (2003), “Comparative review of the longlived wastes transmutation performances in nuclear reactor”, Global 2003 New
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FURTHER READING

Further reading

Canada•Bronwyn Hyland, Gary Dyck, Andrew Morreale and Ragnar Dworschak: Transmutation of
Actinides in CANDU Reactors, 10th Information Exchange Meeting on Partitioning andTransmutation in Mito, Japan on, 6-10 October 2008.

•

B. Hyland and G.R. Dyck: Actinide Burning in CANDU Reactors, Global-2007, Advanced
Nuclear Fuel Cycles and Systems, Boise, ID, September 9 - Friday, September 14, 2007.

•

A.R. Dastur, A.S. Gray, N. Gagnon D.B. Buss, R.A. Verrall: The Role of CANDU in Reducing theRadiotoxicity of Spent Fuel, Global-93, Proceedings: Future of Emerging Nuclear Systems,
Seattle, 1993.

•

J.H. Bultman, N.Gagnon, A.R. Dastur:
Statics and Dynamics of CANDU Transuranics
Burners, International Conference on Evaluation of Emerging Nuclear Fuelcycle Systems,
Global'95, 11-14 September 1995, Versailles, France.

•

P.S.W. Chan, M.J.N. Gagnon, P.G. Boczar, R.J. Ellis, R.A. Verrall: CANDU® – A Versatile
Reactor for Plutonium Disposition or Actinide Burning, Global '97 International Conference
on Future Nuclear Systems, held in Yokohama, Japan, 5-10 October 1997, AECL-11853.

•

T. A. Taiwo, T. K. Kim, F. J. Szakaly, R. N. Hill, and W. S. Yang, G. R. Dyck, B. Hyland,
G.W.R. Edwards: Comparative Study of Plutonium Burning in Heavy and Light Water
Reactors, Proceedings of ICAPP 2007 CW-123700-CONF-002, Nice, France, 13-18 May 2007.

•

P.G. Boczar, M.J.N. Gagnon, P.S.W. Chan, R.J. Ellis, R.A. Verrall, A.R. Dastur: Advanced
CANDU Systems for Plutonium Destruction, NATO Advanced Research Workshop (ARW)on “Advanced Nuclear Systems Consuming Excess Plutonium”, Moscow, Russian Federation.

France•Updates on major recent developments in France, 39th RWMC Meeting 15-17 March 2006.

•

Commission Nationale d’Évaluation Relative aux recherches sur la gestion des dechetsRadioactifs, RÉSUMÉ ET CONCLUSIONS, Rapport d'Evaluation No 11, Juin 2005.

•

Andra research on the geological disposal of high-level long-lived radioactive waste Results
and perspectives, Agence nationale pour la gestion des déchets radioactifs (ANDRA) June
2005.

•

French R&D on Partitioning and Transmutation of Long-lived Radionuclides, AnInternational Peer review of the 2005 CEA Report, ISBN: 978-92-64-03256-9, ISBN: 92-6402296-1, OECD/NEA 2006.

•

J. Journet, G. Vambenepe, J. Vergnes (EdF), F. Biagi, H. Sztark, Minor Actinides
Transmutation in Oxide Fuelled Fast Reactors, Global-93, Proceedings, Future of Emerging
Nuclear Systems, Seattle, 1993.

•

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M. Delpech, J. Tommasi, A. Zaetta, M. Salvatores, H. Mouney, G. Vambenepe:
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F Varaine, C Chabert, G Youinou, D Warin: Transmutation in industrial thermal reactors(PWR), ENC 2005, Versailles, France, 11-14 December 2005•

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J.P. Grouiller, L. Boucher, P. Bourdot, F. Varaine, M. Delpech, D. Warin: A synthesis ofpossible separation and transmutation scenarios studied in the frame of the French law for
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A. Vasile, F. Varaine, M. Delpech, Minor actinides recycling in current pressurized water
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T. Mukaiyama, H. Takano, T. Ogawa et al. “Partitioning and Transmutation Studies at JAERIBoth Under OMEGA Program and High Intensity Proton Accelerator Project”, Progress in
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H. Takano, T. Ikegami, “Activities on R&D of Partitioning and Transmutation in Japan”, Inter.
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H. Takano, H. Oigawa, T. Sasa et al. “Activity on R&D for P&T in JAERI”, Proc. of International
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Kotchetkov, I. Krivitskiy, N. Rabotnov, A. Tsiboulia, S. Iougai: Calculation and Experimental
Studies on Minor Actinides Reactor Transmutation, Fifth OECD/NEA Information Exchange
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T. Greifenkamp, G. Ivan Maldonado, J. Gehin, M. DeHart: Analysis of Minor Actinide Target
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R. A. Wigeland, T.H. Bauer, R.N. Hill, J.A. Stillman: Impact on Geologic Repository Usage
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B. Bradley, J. Ragusa: Plutonium and Minor Actinides Transmutation in Standard PWRs Fuel
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R. Sagdeev, P. Tsvetkov, W. Sadowski, G. Maldonado, A. Kritz, J. Gehin, Y. Gohar, C. Liu,
J. Ragusa, W. Charlton, G. Bateman, M. DeHart: Development of a Joint US-RF Educational
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M. K. Meyer, S.L.Hayes, W. J. Carmack, H. Tsai: THE EBR-II X501 Minor Actinide Burning
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Dale Lancaster: Actinide Burning In A Standard Pressurized Water Reactor, Global-93,
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K. D. Dobbin, J.V. Nelson, S.F. Kessler, R.P. Romberg, D.W. Wootan: Evaluating the Efficacy ofa Minor Actinide Burner: Global-93, Proceedings: Future of Emerging Nuclear Systems,
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R.N. Hill, T.A. Taiwo, J.A. Stillman, D.J. Graziano, D.R. Bennett, H. Trellue, M. Todosow,
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T. Taiwo, F. Szakaly, T-K Kim, R. Hill: “Fuel Cycle Impacts of Uranium-Plutonium Coextraction”Physor-2008, International Conference on the Physics of Reactors “Nuclear
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E.A. Hoffman, R.N. Hill, T.A. Taiwo: “Advanced LWR Multi-Recycle Concepts”, Trans. ANS,
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E. Shwageraus, P. Hejzlar, M.S. Kazimi: “Feasibility of Multirecyling of Pu and MA in
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J.A. Stillman, R.N. Hill: “Effect of Thermal-Spectrum Transmuter Deep Burnup of
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T. A. Taiwo, T. K. Kim, J. A. Stillman, R. N. Hill, M. Salvatores, P. J. Finck: “Assessment of a
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M. Visosky, P. Hejzlar, M. Kazimi, Actinide Transmutation Using Pressurized Water Reactors,
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R.A. Wigeland, T.H. Bauer, T.H. Fanning, E.E. Morris “Repository Impact of LWR MOX andFast Reactor Recycling Options”.

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MINOR ACTINIDE BURNING IN THERMAL REACTORS, NEA No. 6997, © OECD 2013

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Advanced Nuclear Fuel Cycles and Radioactive Waste Management OECD/NEA 2006 ISBN:
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Innovation in Nuclear Energy Technology, © OECD 2007, NEA No. 6103 Nuclear Energy, ISBN:
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F. Ganda, E. Greenspan (University of California, US): Plutonium and minor actinides multirecycling in PWR using hydride fuels, log590•

Letourneau, O. Bringer (CEA, France), I. Al Mahamid (New York State Dep of Health, US),F. Chartier, E. Dupont (CEA, France), P. Mutti (ILL, France), L. Oriol, S. Panebianco, Ch.
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MINOR ACTINIDE BURNING IN THERMAL REACTORS, NEA No. 6997, © OECD 2013

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Minor Actinide Burning in Thermal

Reactors

A modern light water reactor (LWR) of 1 GWe capacity will typically discharge about 20-25 tonnes ofirradiated fuel (spent fuel) per year of operation. Despite the low content of about 0.1-0.2% of minor
actinides in spent fuel, these actinides can nonetheless contribute significantly to decay heat loading andneutron output, as well as to the overall radiotoxic hazard of spent fuel. For this reason, there has long
been an interest in transmuting minor actinides to reduce their impact on the back end of the fuel cycle.
Fast reactors are needed to effectively transmute transuranics (TRUs), including minor actinides. However,recent studies have demonstrated that TRU transmutation rates can also be achieved in thermal reactors,
although with certain limitations due to the accumulation of transuranics through recycling and theirimpact on the safety of power plants. The transmutation of TRUs could potentially be implemented in a
substantial number of thermal reactors operating today, while waiting for a similar programme in fast
reactors to allow for commercial-scale operations in 20 to 30 years or more.
This publication provides an introduction to minor actinide nuclear properties and discusses some of thearguments in favour of minor actinide recycling, as well as the potential role of thermal reactors in thisregard. Various technical issues and challenges are examined from the fuel cycle, operations, fuel designs,
core management and safety/dynamics responses to safety and economics. The focus of this report ison the general conclusions of recent research that could be applied to thermal reactors. Further researchand development needs are also considered, with summaries of findings and recommendations for the
direction of future R&D efforts.

OECD Nuclear Energy Agency
12, boulevard des Îles92130 Issy-les-Moulineaux, France
Tel.: +33 (0)1 45 24 10 15
nea@oecd-nea.org www.oecd-nea.org

NEA No. 6997