U-Mo Monolithic Fuel for Nuclear Research and Test Reactors
U-Mo Monolithic Fuel for Nuclear Research and Test Reactors
RAMPRASHAD PRABHAKARAN 0
0 1.-Pacific Northwest National Laboratory , Richland, WA, USA. 2.- , USA
Research and test reactors consist of a wide range
of civil and commercial nuclear reactors that are
generally not used for power generation. The
primary purpose of these reactors is to provide a
neutron source for research and development
purposes. These reactors are used for a number of
applications, such as testing and analysis of
materials, industrial processing and production of
radioisotopes. In addition to the nuclear field, these
reactors are also used in other areas, such as
physics, chemistry, biology, geology, archeology,
environmental science, and medicine.1,2
As per the International Atomic Energy Agency
(IAEA) database dated April 2016, there are 243
operational research reactors, 7 under construction,
11 being planned, 134 reactors have been
permanently shut down and 352 reactors have been
decommissioned. About half of the operational
research reactors are over 40 years old.3 The U.S.
Nuclear Regulatory Commission (NRC) regulates 42
research and test reactors of which 31 are currently
operating. Most research and test reactors are at
universities or colleges in the United States.2
Research and test reactors are smaller in size and
operate at lower temperature (typical coolant
temperature is below 100 C) when compared to power
reactors, but the operating conditions are more
rigorous. The peak power density is about 5 kW/cc
for a typical power reactor, whereas it could be
about 17 kW/cc (in the fuel meat) for a typical
research and test reactor. The burn-up is also very
high in a research and test reactor. In a power
reactor, burn-up is limited to less the 10% of the
heavy metal while many research reactors will see
complete depletion of heavy metal in peak
locations.4 The power of a typical power reactor is about
3000 MWt (sufficient to power about 200,000
households in the peak summer), whereas it is only in the
range of 0.10 W (sufficient to power a night lamp)
Ramprashad Prabhakaran is the JOM advisor for the Nuclear Materials
Committee of the TMS Structural Materials Division, and guest editor for
the topic U-Mo Monolithic Fuel for Nuclear Research and Test Reactors in
and 20 MWt (sufficient to power about 20 standard
medical x-ray machines) for a typical research and
test reactor.2 These reactors are also covered by
IAEA safety inspections and safeguards, similar to
power reactors. These reactors employ a wider
range of designs when compared to power reactors.5
About 80% of the world’s power plants are classified
into two basic types (pressurized water reactors and
boiling water reactors).3
The first research and test reactors built around
the 1940s which employed LEU (low enriched
uranium: < 20 wt.% U-235) fuel were low-powered
reactors, used mainly for studying reactor physics
and reactor technology. However, due to the
increased use of these reactors for a number of
applications, the demand for higher specific power
and the need to use greater U-235 concentrations
increased, thus leading to the use of HEU (high
enriched uranium: > 20 wt.% U-235; typically, 90%
enriched) fuel, instead of LEU fuel.4
The Reduced Enrichment for Research and Test
Reactors (RERTR) Program was initiated by the
United States Department of Energy in August
1978, in response to the increased concern about the
potential diversion of HEU for use in nuclear
weapons.4,6 Since the 1980s, the United States
policy has encouraged the use of LEU fuels for all
new research and test reactor designs worldwide,
and also for the conversion of the existing reactors
from the HEU to LEU fuel.7
The RERTR program (now called the Reactor
Conversion Program under the National Nuclear
Security Administration’s Office of Material
Management and Minimization) identified 106 research
and test reactors in the United States and overseas
for conversion to LEU fuel.8 Eighty-seven of the
targeted 106 research and test reactors have been or
can be converted from the HEU to LEU fuel, using
the RERTR qualified dispersion fuels (fuel consists
of fuel kernels surrounded by aluminum matrix
material; density: 8 g/cm3).9 The remaining 19
reactors have an exotic geometry and/or are
highpower/high-flux reactors. A majority of the
remaining high-power research and test reactors
still operating on HEU fuel have fissile atom density
requirements that are too high for conversion to
existing LEU fuel elements, i.e., dispersion fuels,
since a decrease in uranium enrichment requires an
increase in the uranium density (> 14.5 g/cm3) to
maintain the net fissile atom density of the
This has led to a new pursuit of developing a high
uranium density monolithic fuel that possesses the
greatest possible uranium density in the fuel region.
Based upon the density requirements and
irradiation performance, metallic uranium alloy was
chosen as a superior candidate for fuel materials.5
Uranium has some material drawbacks, such as
poor oxidation and corrosion resistance, low
hardness and yield strength, and lack of dimensional
stability of the room-temperature alpha phase.
Dimensional stability of the fuel during reactor
operation is extremely important.12 Therefore, the
high-temperature gamma (c) phase is desired, based
on the isotropy that can be retained at room
temperature and better resistance to thermal
recycling and radiation damage.13 In order to stabilize
the high-temperature c phase, an alloying element
is added, such as molybdenum (Mo) that has a high
solid solubility in bcc c uranium.10 The U-Mo phase
diagram shows that U-10Mo is near the eutectoid
transformation of gamma-U and provides a good
compromise between the amount of alloying metal
needed for phase stability and the fuel density.14,15
The metallic fuel selected to replace the current
HEU fuels is the LEU-10wt.% Mo alloy in the form
of a thin sheet or foil encapsulated in AA6061
aluminum alloy with a zirconium interlayer.12,16
The U-10Mo monolithic fuel can achieve the desired
higher uranium density (15.6 g/cm3). In order to
effectively lead this investigation, new
developments in processing and fabrication of the fuel
elements have been initiated, along with a better
understanding of material behavior before and after
irradiation as a result of these new developments.
Complex materials processing techniques such as
casting, thermal annealing, hot and cold rolling,
coating, and hot isostatic pressing are being used to
fabricate LEU-10Mo fuel plates. Hence, it is very
important to obtain
processing–structure–properties correlations. Efforts are ongoing to utilize
computational modeling to understand these
complex thermomechanical processes. The article
written by Xu et al. describes the application of
integrated computational materials engineering
(ICME) concepts to integrate three individual
modeling components (homogenization,
microstructurebased finite element method for hot rolling, and
carbide particle distribution) to simulate the early
stage of LEU-10Mo alloy manufacturing processes.
As a part of LEU-10Mo monolithic fuel
development, reactor experiments are being performed in
the Advanced Test Reactor (Idaho Falls, ID, USA).
To support fuel qualification, the fuel should exhibit
mechanical integrity, geometric stability, and
stable and predictable irradiation behavior at high
powers and high fission densities. The article
authored by Keiser et al. provides an overview of
the microstructures observed at different regions of
interest in fuel plates (fabricated using
laboratoryscale methods) before and after irradiation for the
fuel samples that have been tested. Discussions
regarding observed microstructural changes during
irradiation that may impact fuel performance are
presented in this article.
Efforts are ongoing to perform irradiation testing of
this new fuel system to support generic regulatory
approval that will allow the fuel to be used in
subsequent reactor licensing and conversion efforts.
Post-irradiation non-destructive examination (such
as visual, neutron radiography, profilometry, and
precision gamma scanning) and subsequent analyses
are used to demonstrate that the fuel meets
established irradiation performance requirements for
mechanical integrity, geometric stability, and
stable and predictable behavior. The article written
by Williams et al. presents the results of
post-irradiation non-destructive examination performed on four
curved full-size fuel plates (AFIP-7 experiment) which
were irradiated under moderate operating conditions
in the Advanced Test Reactor in order to evaluate the
fuel performance for geometries that are prototypic of
research reactor fuel assemblies.
Irradiation-induced recrystallization, a general
phenomenon that has been observed in various
nuclear fuels (UO2, U-Mo dispersion and monolithic
fuels) can accelerate inter-granular gas bubble
growth rates and swelling kinetics. The article
authored by Hu et al. presents the results of a
recently developed recrystallization model to study
the effect of microstructures and radiation conditions
on recrystallization kinetics in U-10Mo fuels. This
model integrates the rate theory of intragranular gas
bubbles and interstitial loop evolutions, and a
phasefield model of recrystallization zone evolution.
The following articles are published under the
topic ‘‘U-Mo Monolithic Fuel for Nuclear Research
and Test Reactors’’ in the December 2017 issue (vol.
69, no. 12) of JOM and can be accessed via the JOM
page at http://link.springer.com/journal/11837/69/
‘‘Process modeling early stages of U-10wt.%
molybdenum alloys using integrated
computational materials engineering concepts’’ by Zhijie
Xu, Xiaowo Wang, Ayoub Soulami, Xiaohua Hu,
Curt Lavender, and Vineet Joshi.
‘‘Observed changes in as-fabricated U-10Mo
monolithic fuel microstructures after irradiation
in the Advanced Test Reactor’’ by Dennis Keiser,
Jan-Fong Jue, Brandon Miller, Jian Gan, Adam
Robinson, and James Madden.
‘‘Post-irradiation non-destructive analyses of the
AFIP-7 experiment’’ by Walter Williams, Adam
Robinson, and Barry Rabin.
‘‘A rate-theory-phase-field model of
irradiationinduced recrystallization in UMo nuclear fuels’’
by Shenyang Hu, Vineet Joshi, and Curt
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