ISSN:1369 7021 © Elsevier Ltd 2010DECEMBER 2010 | VOLUME 13 | NUMBER 1214

Materials challenges for nuclear systems The safe and economical operation of any nuclear power system relies to a great extent, on the success of the fuel and the materials of construction. During the lifetime of a nuclear power system which currently can be as long as 60 years, the materials are subject to high temperature, a corrosive environment, and damage from high-energy particles released during fission. The fuel which provides the power for the reactor has a much shorter life but is subject to the same types of harsh environments. This article reviews the environments in which fuels and materials from current and proposed nuclear systems operate and then describes how the creation of the Advanced Test Reactor National Scientific User Facility is allowing researchers from across the United States to test their ideas for improved fuels and materials.

Todd Allena, Jeremy Busbyb, Mitch Meyerc, David Pettic

a Department of Engineering Physics, University of Wisconsin, Madison, WI 53706, USA b Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831, USA c Nuclear Science and Technology Division, Idaho National Laboratory, Idaho Falls, ID, United States

* E-mail: [email protected]

Successful operation of current light water reactors and

implementation of advanced nuclear energy systems is strongly

dependent on the performance of fuels and materials. A typical

Light Water Reactor (LWR) contains numerous types of materials

(Fig. 1) that must all perform successfully. A majority of the

LWRs in the U.S. are extending their operating licenses from a 40

year period to a 60 year period, with initial discussions about 80

year lifetimes now underway. Many proposed advanced systems

(also known as Generation IV systems) anticipate operation at

temperatures and radiation exposures that are beyond current

nuclear industry experience, as well as most previous experience

with developmental systems1-7.

Table 1 summarizes the expected environments during normal

operation for the six Generation IV systems. For comparison, the

operating conditions for a Pressurized Water Reactor (a type of light

water reactor) are also listed. The Generation IV systems are expected

to operate at higher temperatures, to higher radiation doses, at

higher pressures, and in some cases with coolants that present more

challenging corrosion problems than current LWRs. Generation IV

systems are expected to operate for at least 60 years.

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For existing LWRs, extending the lifetime of each fuel element

would improve the energy extraction from the fuel, limit the total

amount of unused fuel (approximately 95% of the energy content

remains at the end of the current useful life of a typical LWR fuel

pin), and improve the overall economics of the plant. For many of the

proposed advanced systems, specifically the fast spectrum systems

like the Sodium Fast Reactor (SFR), Lead Fast Reactor (LFR), and Gas

Fast Reactor (GFR), advanced fuel forms purposefully contain fission

products from previously used fuel with the goal of burning these

fission products to reduce the long-lived radioactivity associated

with the fuel. These fast reactor fuels, in addition to having different

compositions, are exposed to different reactor conditions. Since these

fast reactor fuels are less technologically developed, a test program is

needed to prove the fuels perform as anticipated.

Fig. 1 Outline of PWR Components and Materials. Courtesy of R. Staehle.

Table 1 Approximate operating environments for Gen IV systems

Reactor Type Coolant Inlet Temp (°C)

Coolant Outlet Temp (°C)

Maximum Dose (dpa*) Pressure (Mpa) Coolant

Supercritical Water-cooled Reactor (SCWR) 290 500 15-67 25 Water

Very High Temperature gas-cooled Reactor (VHTR) 600 1000 1-10 7 Helium

Sodium-cooled Fast Reactor (SFR) 370 550 200 0.1 Sodium

Lead-cooled Fast Reactor (LFR) 600 800 200 0.1 Lead

Gas-cooled Fast Reactor (GFR) 450 850 200 7 Helium/ SC CO2

Molten Salt Reactor (MSR) 700 1000 200 0.1 Molten Salt

Pressurized Water Reactor (PWR) 290 320 100 16 Water

* dpa is displacement per atom and refers to a unit that radiation material scientists used to normalize radiation damage across different reactor types. For one dpa, on average each atom has been knocked out of its lattice site once.

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An additional source of uncertainty also exists with extended

operation or new operating regimes: the potential for new forms of

degradation. For example, in the area of radiation effects, in the past,

when new reactor operating conditions (temperature, flux, or fluence)

have been established at least one new radiation-induced phenomenon

has been found. In the 1960s irradiation-induced hardening was

discovered. Swelling was a major concern for fast reactors in the 1970s

and high-temperature embrittlement due to helium was a surprise in

the 1980s. For new Generation IV systems or the extension of current

technology, one should be aware of the possibility of new phenomena

due to irradiation, corrosion, or aging in both materials and fuels

performance.

Because of the challenges to fuels and materials in both currently

operating LWRs, as well as the proposed advanced systems, facilities

for testing fuels and materials are critical. The Department of Energy

opened the Advanced Test Reactor (ATR) at the Idaho National

Laboratory as a user facility in 2007, allowing access to reactor test

space and post-irradiation examination facilities through an open

solicitation and project selection based on peer review. The ATR

National Scientific User Facility (ATR NSUF) now provides the nuclear

energy research community a means of testing concepts with the

potential to improve the ability of current and advanced nuclear

systems to benefit operating performance, economics, safety, and

reliability.

Many countries across the world are working on advanced reactor

concepts and while each may use materials with a unique designation

system, the fuels and materials used are typically similar and the

challenges outlined in this article are common, whether the researcher

is from Europe, India, Japan, South Korea, Russia, the United States or

any of the other countries researching fuels or materials for nuclear

systems. This review article outlines some of the challenges associated

with materials and fuels for nuclear systems and describes the ATR

NSUF.

Challenges for materials in nuclear power systems Nuclear reactors present a harsh environment for component service

regardless of the type of reactor. Components within a reactor core

must tolerate exposure to the coolant (high temperature water, liquid

metals, gas, or liquid salts), stress, vibration, an intense field of high-

energy neutrons, or gradients in temperature. Degradation of materials

in this environment can lead to reduced performance, and in some

cases, sudden failure.

Materials degradation in a nuclear power plant is extremely complex

due to the various materials, environmental conditions, and stress

states. For example, in a modern light water reactor, there are over

25 different metal alloys within the primary and secondary systems

(Fig. 1); additional materials exist in concrete, the containment vessel,

instrumentation and control equipment, cabling, buried piping, and

other support facilities. Dominant forms of degradation may vary

greatly between different systems, structures, and components in

the reactor and can have an important role in the safe and efficient

operation of a nuclear power plant. When this diverse set of materials

is placed in the reactor environment, over an extended lifetime,

accurately estimating the changing material behaviors and service

lifetimes becomes complicated.

Today’s fleet of power-producing light water reactors faces a very

diverse set of material challenges. For example, core internal structures

and supports are subjected to both coolant chemistry and irradiation

effects. These stainless steel structures may experience irradiation-

induced hardening, radiation-induced segregation and changes to the

microstructure. In addition, these factors may lead to susceptibility to

irradiation-assisted stress corrosion cracking as shown for a baffle bolt

in Fig. 2.

The reactor pressure vessel, a low-alloy steel component, also

experiences radiation-induced changes and can be susceptible

to embrittlement. The last few decades have seen remarkable

progress in developing a mechanistic understanding of irradiation

embrittlement7. This understanding has been exploited in formulating

Fig. 2 Examples of stress-corrosion cracking in LWR power plants. (a) Primary water stress corrosion cracking in steam-generator tubing and (b) irradiation- assisted stress corrosion cracking in a PWR baffle bolt.

(b)

(a)

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robust, physically-based and statistically-calibrated models of Charpy

V-notch (CVN)-indexed transition temperature shifts. The progress

notwithstanding, however, there are still significant technical issues

that need to be addressed to reduce the uncertainties in regulatory

application.

Components in the secondary (steam generator) side of a nuclear

reactor power plant are also subject to degradation. While the

secondary side of the reactor does not have the added complications of

an intense neutron irradiation field, the combined action of corrosion

and stress can create many different forms of failure. The majority

steam generator systems in US power plants today originally used

Alloy 600 (a Ni-Cr-Fe alloy), although service experience showed

many failures in tubes through the 1970s. In the last 20 years, most

steam generators have been replaced with Alloy 690, which shows

more resistance to stress-corrosion cracking. In addition to the base

material, there are weldments, joints, and varying water chemistry

conditions leading to a very complex component. Stress-corrosion

cracking is found in several different forms and may be the limiting

factor for component lifetime. The integrity of these components is

critical for reliable power generation in extended lifetimes, and as a

result, understanding and mitigating these forms of degradation is very

important.

In general, concrete structures can also suffer undesirable

changes with time because of improper specifications, a violation of

specifications, or adverse performance of its cement paste matrix or

aggregate constituents under environmental influences (e.g., physical

or chemical attack). Some examples are shown in Fig. 38-11. Changes

to embedded steel reinforcement as well as its interaction with

concrete can also be detrimental to concrete’s service life. A number

of areas of research are needed to assure the long-term integrity

of the reactor concrete structures. A database with a compilation

of performance data under service conditions is an initial need. An

additional requirement is a systematic and mechanistic understanding

of the mechanical performance impacts from the long-term effects of

elevated temperature and, for some locations, the effects of irradiation.

In addition to LWR technology, a broad variety of advanced reactor

systems are currently being considered and developed in the United

States. For example, the Generation IV programs are examining

reactors ranging from sodium fast reactors to gas-cooled reactors to

liquid salt-cooled reactors. This breadth of designs creates a great

range in operating conditions for materials (Table 1). For example,

core internal structures must tolerate sodium at 500 °C to ~10 dpa

while fuel cladding and duct materials may be required to survive up

to 200 dpa in the same coolant4. Components in high temperature

gas reactors may reach temperatures up to 1000 °C while liquid salt

reactors may require even higher temperatures. Lead or lead alloys

provide excellent heat transfer leading to inherently safe reactors but

typical construction materials made of Fe, Cr, and Ni are soluble in

lead so specific high-temperature corrosion protection methods need

to be devised to take full advantage of these coolants. Molten salts

Fig. 3 Examples of degradation in concrete structures. Courtesy of D. Naus.

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provide similar heat transfer characteristics to water but would not

have to be pressurized, leading to increased safety under a pipe break.

The challenge for many candidate molten salts is that they do not form

protective oxides with steels, making corrosion protection the critical

issue also. These extreme environments demand advanced materials for

successful service.

Advanced materials have the potential to improve reactor

performance via increased safety margins, design flexibility, and

economics and overcome current reactor performance limitations.

Increased strength and creep resistance can give greater design margins

leading to improved safety margins, longer lifetimes, and higher

operating temperatures, thus enabling greater flexibility. Improved

mechanical performance may also help reduce the plant capital cost

for new reactors both by reducing the required commodities (with

concomitant reductions in welding, quality assurance and fabrication

costs) and through design simplifications. Successful implementation,

however, requires considerable development and licensing effort.

Modern materials science tools such as computational thermodynamics

and multi-scale radiation damage models, in conjunction with rapid

science-guided experimental validation, may offer the potential for a

dramatic reduction in the time period to develop and qualify structural

materials.

There are many requirements for all nuclear reactor structural

materials, regardless of the exact design or purpose. The material must

have adequate availability, fabrication and joining properties, as well as

favorable neutronic and thermal properties. Further, it must have good

mechanical properties, good creep resistance and long-term stability.

Sufficient data under the range of in-core operating conditions must be

available to support the licensing process. Finally, since the materials

will be used in a high-energy and intensity neutron field, it must be

tolerant of radiation effects. When selecting structural materials for

any fission reactor application, a careful trade-off analysis is needed for

each specific reactor design. Reactor characteristics including operating

temperature, coolant, neutron flux, neutron spectrum, fuel type, and

lifetime must also be considered to select the most suitable structural

material.

Another common need regardless of the advanced reactor design

being considered is a detailed understanding of compatibility issues

between the structural material and the coolant. Compatibility

between the structural materials and coolant is a vital consideration

in any reactor design process. The coolant selection is based on the

required thermal properties, such as low melting point, high heat

transfer coefficient, etc., and the expectation that structural and clad

materials are generally compatible with the coolant (regardless of if

it is water, liquid metals, or molten salts) in terms of corrosion and

chemical interactions. Today, the most mature fast reactor designs

are all sodium cooled fast reactors. While there is considerable

experience with this coolant in fast reactor applications in the U.S. and

internationally, there is little recent experience in sodium compatibility

and only scarce data on new alloys currently being developed.

Only through careful evaluation of all factors and a thorough trade

analysis will the most promising candidate materials be chosen for

further development. It is important to note that there is no ideal

material that is best for each of the considerations listed. Indeed,

all candidate materials have advantages and limitations. The most

promising alloys, which allow the best performance, are also the least

technically mature and will require the most substantial effort. These

trade-offs must all be weighed carefully.

A systematic and science-based approach can reduce both time and

expense required for development, validation, and qualification. This

approach may also enable improvements in performance by optimizing

alloy composition and processing for specific service conditions. Using

a combination of computational tools and more advanced analytical

techniques will greatly accelerate research over past advanced reactor

material development programs.

Challenges in the development of nuclear fuels Nuclear reactors are built around a core of fuel. The performance of

reactor systems is determined by the performance of the fuel. The

inherent physical features of the fuel, such as thermal conductivity,

diffusion rates of gaseous species, and chemical compatibility of

the fuel and cladding, in turn, determine the performance of the

fuel system. Enabling significant improvements in nuclear reactor

and nuclear fuel cycle technology depends, to a large degree, on the

understanding and development of robust new fuel systems.

The development of nuclear fuel presents many technical

challenges. In-reactor fuel behavior is complex, affected by steep

temperature gradients and changes in fuel chemistry and physical

properties that result from nuclear fission. These challenges are

compounded by the highly radioactive nature of irradiated fuel, and

the necessity of conducting fuel examinations remotely, in a heavily

shielded environment.

Light water reactor fuel challenges The majority of the world’s commercial nuclear power plants are light

water reactors. These reactors, after more than 50 years of operational

experience, have proven to be extremely successful, generating

emission free electricity at a cost competitive with that of coal-fired

plants. Worldwide, 359 LWRs operate with a generating capacity of

338 GWe; LWR plants produce 87% of all nuclear electricity and a total

of 14% of the world’s total electricity12.

Current commercial LWRs use a core of zirconium alloy clad UO2

fuel (Fig. 4). Since the 1990s, average fuel burn-up (burn-up is a term

describing the fuel’s lifetime) has nearly doubled, power uprates of

existing plants in the United States have resulted in an increase in

energy output equivalent to 27 new nuclear plants since 1973, and

cycle lengths have increased. Mitigation of stress corrosion cracking of

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plant materials with chemistry additions and fuel loading that result

in low neutron leakage has also occurred. These changes in operation

have resulted in steady increases in power production, but also placed

additional stress on the fuel. Fuel failures are not due to the failure of

the fissile material, but of the cladding that encapsulates the fuel and

separates it from the reactor coolant. Fuel failures, while not significant

to plant safety, negatively affect the economics of nuclear plant

operation, often requiring plant power restrictions or plant shutdown

to replace the leaking assembly. These failures have been aggressively

managed by the nuclear industry13. Approximately 70% of fuel LWR

failures are caused by vibration induced wear and cladding penetration

by foreign matter14. The remaining 30% of failures are due to CRUD

deposits, pellet cladding interaction, and unknown or unassigned

causes. CRUD is a tenacious iron, nickel, chromium oxide deposit that

forms as the result of deposition of stainless steel corrosion products

on the fuel surface which results in altered heat transfer from the fuel.

Pellet cladding interaction failures initiate during fuel power changes

at locations where there are defects in fuel pellet surfaces due to a

combination of fission product attack and stress concentration. Also of

concern, if a loss of coolant event occurred, is hydrogen uptake by the

zirconium alloy cladding, which can lead to cladding embrittlement.

Given the adverse consequences of fuel failure and commercial

limits on uranium enrichment, the practical burn-up limit of current

LWR fuels is likely to be in the range of 65 – 75 GWd/MTU15. It may

be possible to progress beyond this range, either through continued

incremental improvements in current fuel technology or by adoption of

advanced fuels. Improvements in current fuels would require addressing

the primary fuel failure modes discussed above, as well as additional

issues that arise at higher burn-up. These additional issues include

accelerated irradiation growth of zirconium alloys, management of

additional fission gas inventory, the degradation of the mechanical

properties of the zirconium cladding with increased radiation damage,

corrosion and hydrogen uptake, and the impact of zirconium alloy

property changes and increased fission gas inventory on fuel behavior

during Reactivity Initiated Accidents16 (RIA) and Loss Of Coolant

Accidents (LOCA)17.

An alternate course of action is the development of robust new

fuels. These fuels include advanced cladding concepts such as silicon

carbide18,19, liquid metal bonded hydride fuel20, high conductivity

metallic fuel, and composite fuels21,22. Concepts such as these

offer potentially large performance benefits, but may require costly

changes to the installed nuclear infrastructure, such as those required

for increased enrichment. Advanced fuels will also be required to

undergo a long and rigorous licensing process. Based on these factors,

deployment of advanced LWR fuels may be possible in the 10-20 year

time frame. The journey to deployment of advanced fuels begins with

irradiation testing23 of fuels concepts that have been the subject of

careful systems analysis to establish feasibility from a fuel performance

perspective.

Beyond electricity: Fuels of high temperature reactors High Temperature Gas-cooled Reactors (HTGRs) are graphite-

moderated nuclear reactors cooled by helium. The high outlet

temperatures and high thermal-energy conversion efficiency of HTGRs

enable an efficient and cost-effective integration with non-electricity

generation applications, such as process heat and/or hydrogen

production, for the many petrochemical and other industrial processes

that require temperatures between 300 °C and 900 °C. Using HTGRs

in this way would supplant the use of premium fossil fuels, such as

oil and natural gas, improve overall energy security in the U.S. by

reducing dependence on foreign fuels, and reduce CO2 emissions. Key

characteristics of this reactor design are the use of helium as a coolant,

graphite as the moderator of neutrons, and ceramic-coated particles

Fig. 4 Schematic of a light water reactor fuel rod.

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as fuel. Helium is chemically inert and neutronically transparent. The

graphite core slows down the neutrons and provides high-temperature

strength and structural stability for the core and a substantial heat

sink during transient conditions. The ceramic-coated particle fuel is

extremely robust and retains the radioactive byproducts of the fission

reaction under both normal and off-normal conditions.

The TRISO-coated (TRIstructural-ISOtropic) particle fuel forms the

heart of the HTGR concept. Such fuels have been studied extensively

over the past four decades around the world including in the United

Kingdom, Germany, Japan, the United States, Russia, China, and more

recently South Africa24. As shown in Fig. 5, the TRISO-coated particle

is a spherical-layered composite, about 1 mm in diameter. It consists

of a kernel of uranium dioxide (UO2) or uranium oxycarbide (UCO)

surrounded by a porous graphite buffer layer that absorbs radiation

damage and allows space for fission gases produced during irradiation.

Surrounding the buffer layer is a layer of dense pyrolytic carbon called

the Inner Pyrolytic Carbon layer (IPyC), a silicon carbide (SiC) layer,

and a dense Outer Pyrolytic Carbon layer (OPyC). The pyrolytic carbon

layers shrink under irradiation and create compressive forces that

act to protect the SiC layer, which is the primary pressure boundary

for the microsphere. This three-layer system is used to both provide

thermomechanical strength to the fuel and contain fission products.

An HTGR will contain billions of TRISO-coated particles encased in a

graphitic matrix in the form of either small cylinders, called compacts,

or tennis-ball-sized spheres, called pebbles (see Fig. 5). Extensive

testing has demonstrated the outstanding performance of high-quality

low-defect TRISO-coated particle fuels. In the German program in the

1970s and 1980s, over 400 000 TRISO-coated UO2 particles were

irradiated to burn-ups of about 9% at temperatures between 1100 °C

and 1150 °C without any failures. Similar results on somewhat smaller

particle populations have been obtained with Japanese and Chinese

fuels irradiated to low burn-up. About 300 000 TRISO-coated UCO

particles have recently completed irradiation in the United States,

and no failures have occurred at a peak temperature of 1250 °C up

to a peak burn-up of 19%25. Testing of German fuel under simulated

accident conditions in the 1980s has showed similarly excellent

performance. Tests of more than 200,000 irradiated TRISO-coated UO2

particles in both pebble and compact fuel forms have demonstrated

Fig. 5 High temperature gas reactor fuel system, showing TRISO fuel particles consolidated into a graphite matrix as prismatic blocks (upper right) or pebbles (lower right).

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no particle failure after hundreds of hours at 1600 °C and significant

retention of important fission products in the fuel element26. Similar

testing has just begun in the U.S. It is this performance, combined with

the passive safety features of modern modular HTGRs, that allows

these reactors to be located in close proximity to industrial complexes

where they can provide heat for high-temperature chemical processes

needed for hydrogen production, chemical synthesis, and petrochemical

industries.

Significant research and development related to TRISO-coated

fuels is underway worldwide. The fuel system is fairly mature and

the current challenge is largely focused on extending the capabilities

of the TRISO-coated fuel system for higher burn-ups (10–20%) and higher operating temperatures (1250 °C) to improve the attractiveness

of high-temperature gas-cooled reactors as a heat source for large

industrial complexes where gas outlet temperatures of the reactor

would approach 950 °C27. Of greatest concern is the influence of

higher fuel temperatures and burn-ups on fission product interactions

with the SiC layer leading to degradation of the fuel and the release

of fission products. Activities are also underway around the world to

examine modern recycling techniques for this fuel and to understand

the ability of gas reactors to burn minor actinides.

Closing the cycle: Fuels for transmutation The total mass of spent fuel generated from nuclear power production

in Light Water Reactors is relatively small; approximately 30 tons

per 1000 MW electric generating capacity per year. Of this mass,

approximately 96% is uranium and an additional 3% are short-lived

or stable fission products that do not pose major disposal challenges.

Approximately 1 wt.% is composed of transuranic elements; plutonium

(0.9%) and minor actinides (0.1%) that pose challenges for disposal.

The minor actinides include neptunium, americium, and curium.

Among the possible methods proposed for management of

plutonium and the minor actinides is neutron induced fission, during

which less problematic fission product elements are formed by the

‘splitting’ of the heavy transuranic atoms. Neutron transmutation

systems also produce fission energy that can be converted into usable

electricity or process heat. Implementation of neutron transmutation

typically utilizes specialized fuels or targets with high minor actinide

content in either a purpose designed minor actinide burner or a more

conventional reactor system. Suitable in-reactor performance of these

fuels and targets is critical to the operation of neutron transmutation

systems. A cross-sectional micrograph of an irradiated minor actinide

fuel rod tested in the U.S. is shown in Fig. 6.

The fundamental knowledge base of chemical and physical

properties of actinide-bearing materials is limited. This includes details

of phase equilibrium of multi-component system, fuel microstructure

up to reactor operating temperatures; thermophysical properties such

thermal conductivity, heat capacity, and thermal expansion coefficients;

and chemical properties such as the nature and kinetics of reactions

between fuel and cladding material. It is important to understand these

properties and kinetic parameters in order to ensure that fuels designed

for transmutation meet performance criteria. Modeling of the chemical

and physical behavior of these materials is complicated by the presence

of the 5f outer shell electrons, and elucidation of properties and

parameters has relied heavily on empirical studies. Empirical studies,

however, are also complicated by difficulties related to the high activity

of these materials. To guide fuel design, continued work utilizing both

experimental measurements and computational modeling will be

required to provide an understanding of the adequate thermophysical

properties.

To provide the highest long-term benefit to reducing radiotoxicity,

the quantity of minor actinides placed in a repository should be

minimized. The highest potential for material loss occurs during

fuel processing to separate the minor actinides from spent fuel and

during fuel fabrication. Extending fuel burn-up lifetime, thereby

reducing the number of fuel processing cycles is one method of

reducing these fabrication losses. The primary candidate fuels for

minor actinide transmutation are metal alloy and oxide-based

fuels. There is an extensive database on the behavior of plutonium-

bearing (without minor actinides) primary candidate metallic and

oxide fast reactor fuels under irradiation. Both have demonstrated the

capability to achieve burn-up on the order of 200 GWd/MTHM28-30.

The primary barrier to achieving higher burn-up is the strength and

reliability of cladding materials at high radiation dose. In the case of

transmutation fuels, this is exacerbated by high helium gas generation

rates resulting from neutron capture and decay of 241Am that cause

Fig. 6 Cross-sectional micrograph of U-29Pu-4Am-2Np-30Zr metal alloy transmutation test fuel after irradiation to 8.9 x 1020 f/cm3 (∼6 at.% Pu burn-up) in the Advanced Test Reactor.

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fuel behavior to markedly differ from conventional commercial nuclear

fuels31,32.

Helium production is likely to be the most important fuel design

consideration for transmutation scenarios with high minor actinide

content. Helium generation is principally due to neutron capture by 241Am to 242Cm and subsequent alpha decay of the 242Cm to 238Pu.

A rule-of-thumb for estimating helium production from americium is

50 ml He per gram of transmuted 241Am. The wide range of possible

fuel compositions leads to a wide range in the potential for total

helium production. There are two possible approaches to dealing with

helium. The first is to design and operate the fuel under conditions

that promote helium release from the fuel phase to a gas plenum. The

second is to design the fuel to effectively retain fission gas and helium

while maintaining an acceptable level of gas-driven swelling. Two

experiments on americium-bearing oxide fuel effectively demonstrate

these divergent approaches. The SUPERFACT experiment33, which

tested two uranium oxide matrix pins containing 20 wt.% americium

in the Phénix fast spectrum reactor exhibited gas release of <60%,

typical of oxide fast reactor fuel and an acceptable level of fuel swelling

at 4.5 at.% burn-up. The EFTTRA-T4 test34 used a microdispersion of

americium oxide in a magnesium-aluminate spinel matrix. Gas release

was a fraction of that measured in the SUPERFACT pins. Pellets in

this test exhibited volumetric swelling of <18 vol.%, and resulted in

excessive cladding strain. It is clear that if gas is to be retained by the

fuel, the fuel must be designed to account for large amounts of swelling.

The advanced test reactor national scientific user facility: A model for research collaborations The ultimate performance of a fuel or material in a nuclear system

is determined through in-reactor testing. Thus, the availability of

test reactors, hot cells, and examination equipment that can handle

radioactive materials is required to prove the principle of any

advanced concept. The Advanced Test Reactor (ATR) has been in

operation since 1967 and mainly used to support U.S. Department of

Energy (U.S. DOE) materials and fuels research programs. Irradiation

capabilities of the ATR and post-irradiation examination capabilities of

the Idaho National Laboratory (INL) were generally not being utilized

by universities and other potential users due largely the high cost of

using these facilities relative to typical research grant awards. While

materials and fuels testing programs using the ATR continue to be

needed for U.S. DOE programs such as the Fuel Cycle Research and

Development Program and Reactor Concept Research Development, &

Deployment programs, the U.S. DOE recognized there was a national

need to make these capabilities available to a broader user base.

In April 2007, the U.S. DOE designated the Advanced Test Reactor

a National Scientific User Facility (NSUF). As an NSUF, the services

associated with university-led experiment irradiation and post-

irradiation examinations are provided free-of-charge. The U.S. DOE is

providing these services to support U.S. leadership in nuclear science,

technology, and education and to encourage active university/industry/

laboratory collaboration. In the initial concept of the ATR NSUF, the

post-irradiation examination equipment was that available at the INL

hot cells, analytical chemistry laboratories, and electron microscopy

laboratory.

Since opening up the ATR, the NSUF has expanded the ways

potential users can interact with the facility. Specifically:

• Additional capability at INL was opened to potential users,

specifically the ATR Critical Facility which is a low power,

geometrically identical version of ATR which can be used to test

radiation detection systems or validate reactor neutronics codes.

• The post-irradiation examination capability at the INL has been

significantly upgraded with the addition of analytical equipment

such as an electron microprobe, a field emission gun scanning

transmission electron microscope, an atom probe, a scanning Raman

Fig. 7 University partners of the ATR NSUF who are part of an integral national irradiation and post-irradiation testing capability. (b) Universities leading projects currently being supported at the ATR NSUF.

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Materials challenges for nuclear systems REVIEW

DECEMBER 2010 | VOLUME 13 | NUMBER 12 23

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system, an atomic force microscope, and dual beam focused ion

beam systems.

• A sample library was created that allows potential users to propose specific experiments against materials and fuels previously irradiated

in other DOE or industry programs

• Connections are being made with other national user facilities such as the Advanced Photon Source at Argonne National Laboratory

that allows experimenters to analyze samples that are transported

to these complimentary user facilities.

• A network of university partners has joined the NSUF, providing additional capability for irradiation and post-irradiation testing

(Fig. 7 (a) indicates the partner facilities).

The addition of these partner facilities, along with the ties to other

major user facilities, has transformed the NSUF into a distributed

network of facilities that uses the best national capability to support

the best scientific ideas proposed across the nation.

The first full year of implementing the user facility concept was

2008 and since that time the NSUF has initiated work on 23 user-

proposed projects. These projects are listed in Table 2. The projects

break into two classes of experiment:

• New reactor-based projects. These projects involve designing and inserting new fuels and materials into either the ATR or the MITR or

using the ATR Critical Facility. These proposing institutions for these

projects are shown in Fig. 7 (b).

• Post-irradiation only projects. These projects analyze previously irradiated material that is held in the sample library against which

potential users can propose examinations.

Continuing improvements in nuclear energy technology rely on the

development of improved materials and fuels for advanced reactor

systems. Continued life extension of current LWR plants relies on a

thorough understanding of the effects of the reactor environment on

long-term material degradation. Both of these research areas require

access to a specialized nuclear research infrastructure, including high

flux test reactors, radiation shielded research laboratories, and high-

end materials characterization tools dedicated for use on radioactive

materials. The establishment of the ATR NSUF has provided an

effective mechanism for research teams to access this specialized

infrastructure for testing advanced materials and fuel concepts and

for better understanding the degradation of materials and fuels in the

existing reactor fleet.

Acknowledgements Work supported by the U.S. Department of Energy, Office of Nuclear

Energy, under DOE Idaho Operations Office Contract DE-AC07-

05ID14517.

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