RADIATION DAMAGE AND NUCLEAR FUELS
TERM PROJECT
QUANTITATIVE CRITERIA
FOR ACCIDENT TOLERANT FUEL
MARCH 22, 2018
Name – 10000xxxxx
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Contents
1 INTRODUCTION ............................................................................................3
2 CURRENT STATUS .......................................................................................3
3 QUANTITATIVE CRITERIA FOR ACCIDENT TOLERANT FUEL ....6
4 SUMMARY ......................................................................................................8
5 REFERENCES .................................................................................................9
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1
Introduction
Nuclear fuel is the fissile material used in nuclear power plants to start and sustain
nuclear reactions, and to generate power. Some common types of nuclear fuels are
radioisotopes of uranium and plutonium, such as U235 and P239. Different types of
fuels available are oxide fuels, metal fuels, ceramic fuels, liquid fuels, physical forms,
less common types, spent nuclear fuels, radioisotope decay fuels, and fusion fuels.
These fuels are characterized by different melting temperatures, densities, and the
extent of nuclear flux they create. Nuclear fuel is made by filling zirconium tubes
with a mixture of plutonium and natural uranium, or filling the tubes with enriched
uranium pellets. Other fuels such as uranium dioxide is made of uranyl nitrate, mixed
with ammonia and pressed into pellets. There are many other ways in which uranium
fuel is prepared and the fuel depends on the type of nuclear reactors. Some categories
of nuclear reactors are boiling water reactor, pressurized heavy water reactor, gascooled reactor, light water graphite reactor, fast neutron reactor, and others [1].
Irrespective of the type of nuclear reactors, these nuclear fuels have inherent dangers
all along the nuclear life cycle made of phases such as ore extraction, refining,
processing, casting, use in reactors, storage, transportation, and disposal of the spent
fuel. The fuel is highly radioactive and any leaks of the coolant used in the reactors,
can render the area for many kilometers from the leak unsafe. The process of nuclear
fission can become unstable, leading to a breakdown of the core causing rapid fallout
of radioactive material. Transportation of the spent fuel is difficult since any spills
from the container can harm the area where the fissile material falls. Disposal
methods are very dangerous since the spent fuel is dumped in deep salt mines, where
it continues to pollute the area. There is also an opportunity for rogue nations to divert
the spent uranium to make bombs. Any accidents or spills can lead to a radioactive
contamination of the area, which has to be abandoned, and this happened in the
accidents at Chernobyl, Russia, and in Fukushima, Japan. During the reactor
operations, control rods and cooling water along with moderators, control the chain
reaction. Any mistiming or wrong step can lead to a continuous chain reaction that
would never stop, since the half life of uranium and plutonium is 4.468 x 109 years [2].
The alternate solution to the unstable standard nuclear fuel is Accident Tolerant Fuel
(ATF). This fuel has many characteristics and composition that can overcome some of
the dangers associated with nuclear reactor accidents [3]. This paper examines details
of ATF, the current status, and presents quantitative criteria for ATF.
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Current status
ATFs have gained importance after the Fukushima accident and it became clear that
the accident was caused by a quick hydrogen explosion that released radionuclides in
weaker nuclear fuel cladding. Steam in nuclear reactors is under very high pressure
and temperature, and this increases the rate of oxidation of UO2-zirconium alloys,
used in cladding material for uranium pellets. Weakening and rupture of the cladding
cause swelling and ballooning of the cladding material, releasing hydrogen that
explodes. Therefore, fuel claddings have to maintain their performance characteristics
during accident conditions. This has led to the development of ATFs. In ATF systems,
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new materials are used that can tolerate loss of coolant for a longer time while
showing improved performance during normal operations. Therefore, ATF has two
concepts, fuel pellets or fissile material and the cladding over the fuel. The desired
material for fuel cladding should have enhanced thermal conductivity and retention of
fission products. Increasing the thickness of fuel cladding is not possible, since it
would accommodate lesser fuel pellets, and reduce the power generation capacity [3].
This section examines the current status of ATF development in different countries.
Research is underway at General Atomics, US to study Silicon Carbide (SiC) as a
suitable material for ATF. SiC has high temperature strength and it shows high
stability under radiation and oxidation, when compared to zirconium alloys. SiC is
used as a cutting tool to machine steel at very high cutting parameters. The
researchers used SiC and SiC composites for monolithic cladding to provide a tough
and hermitically sealed container for improved performance and safety. While initial
results are promising, challenges are present in fabricating long and thin walled tubes
that meet the functional specifications. The mechanical and thermal characteristics of
unirridated SiC structures were examined along with dimensional control and
permeability. The conclusions indicate that while the material was structurally sound,
it was brittle, there were problems in maintaining tubular geometry, and development
is needed to design and research several areas of SiC [4].
Researchers in Korea studied the feasibility of surface-coated Zr cladding with metalceramic hybrid cladding. The objective was to suppress production of hydrogen
during accidents. Another objective was the development of microcell UO2 pellets to
increase retention of radioactive and corrosive substances such as Cesium and I. The
concept was to cover the UO2 grains in thin cell walls that would trap the movement
of fission products and act as barriers. The research focused on reducing hydrogen
production by steam oxidation at high temperatures so that explosions could be
prevented. Another focus area was to increase the properties of fission products to
retain Cs and I. Two concepts were used in the design. The first was surface
modification by using tubes of Zr alloys with external coating, and the second was to
have Zr alloys tubes with ceramic composite coating and metal-ceramic hybrid
cladding over this. Samples were coated with Cr, Si and Al3Ti alloys, and were
subjected to high temperature steam oxidation. The results showed that samples with
Al3Ti alloys showed very low oxidation at elevated temperatures. Further
developments are to manufacture a hybrid cladding with three parts, a Zr inner liner, a
composite layer made of Zr, and a surface coating on the outside. This construction
allows the matrix to be sealed tight even if the outer later develops cracks. The outer
ceramic later has a higher safety margin since it can withstand very high temperatures.
Challenges remain in developing the ceramic layer [5].
A research was conducted at the Idaho National Laboratory, US to use uranium
silicide pellets as ATF to replace uranium oxide pellets. Powder metallurgy was used
to produce uranium silicide and these pellets had the density greater than the
theoretical density of 94%. Uranium and silicon form different types of stoichiometric
compounds such as USi1, USi2, U3Si2, and others. Among these U3Si2 has higher
uranium silicide content and they can be used to replace UO2. It does not swell under
irradiation and is suitable for use as nuclear fuel. U3Si2 has 17 more atoms than UO2
and this gives superior uranium loading and allows cladding materials such as SiC.
The operating temperature is lower and the thermal conductivity is higher, and the
lower temperature is useful in nuclear reactors. The researchers fabricated the fuel in
lab conditions in an inert glove box since uranium is pyrophoric. Uranium silicide was
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made by combing uranium and silicon powders and a hydride/ dehydride process was
used to produce the powder. The resulting powder was compacted and heated in a
furnace at 1450 degree centigrade and then in an arc smelter, where tungsten electrode
was used to pass an electric current to produce uranium silicide with 97% purity. The
samples were turned into pellets and subjected to different tests such as Scanning
Electron Microscope scanning, chemical and micro hardness analysis, and ATF-1
irradiation. The last step helped to understand the irradiation performance and to
obtain data on possible ATF use. The pellets were pressed into Zr alloy tubes and sent
to INL Hot Fuels Examination Facility to run tests that replicate accident like
conditions in a reactor. Further research is underway to understand the behavior and
swelling characteristics of the pellets and their possible use in reactors as ATF would
depend on the test outcomes [6].
While ATFs are a good concept, several factors need to be considered before standard
fuels are replaced. These factors include backward compatibility, economics and costs,
impact on the fuel cycle and plant operations, and on the plant safety with design
basis and beyond design basis. To develop ATF, researchers selected fully ceramic
micro-encapsulated (FCM) fuel. This idea was initially used in high temperature gas
cooled reactors where the particles were dispersed in a graphite matrix. This concept
was used to develop cladding for LWR fuel with a SiC matrix that enhances stability
under irradiation, acts as a barrier for release of fission products, and resists
proliferation. Cladding materials used are stainless steel of 316 grade, FeCrAl, and
SiC. Tests were run for the fuel, UO2 and Zr alloy cladding. Simulations of accidents
were performed with the MELCOR severe accident code, to evaluate the different
cladding performance. Results showed that the standard UO2-Zr system faced quick
oxidation when the reactor water level fell, releasing hydrogen and heat within 4
hours. The new cladding system of UO2-FeCrAl on the other hand, remained steady
and released half the amount of hydrogen after 8.5 hours. Similar results were seen for
FCM-FeCrAl combination, and the core remained almost intact after relocation.
These effects are due to the high melting temperature of the new cladding. In the
simulation, the UO2-Cr system shows rapid heating of upper structures and it
relocates to the bottom, releasing more hydrogen and steam. For the new cladding,
since cladding oxidation did not take place, the core remained intact. The researchers
conclude that simulations indicate that the UO2-FeCrAl and FCM-FeCrAl systems
can be considered for reactors. However, other factors such as cost, economics,
compatibility, and reliability need to be examined [7].
The previous sections presented research on ATF and several materials and methods
were tested. The physical preparation of samples and preparing tests appears to be a
tedious task. If the research does not meet the objectives, then the whole set of
experiments is wasted. UK is at the forefront of nuclear reactor research and it has set
up the Integral Inherently Safe LWR design (I2S-LWR), 2850 MW integral PWR, for
research and testing. Researchers in UK have developed a lab simulation where it is
possible to create simulations using different lab code such as ANSWER reactor
physics code, WIMS and EDF energy core simulator PANTHER. The objective of the
research was to study the performance of uranium nitride and cladding material
FeCrAl and silicon carbide. In the study, I2S-LWR was designed with U3Si2 and
stainless steel. The reactor has an integrated feature that allows cooling to be removed
and simulate accidents. In the experiment, the reactor was tested for the fuel and steel
alloy cladding using a two stage approach. Macroscopic cross sections were simulated
with WIMS and used for a 3D core design with PANTHER. The results from this
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simulation were compared with Monte Carlo code and the results examined. The
results indicate that there was a high co-relation between the reactor physics code and
the WIMS code, showing that the simulations have a high reliability level. This study
is important since researchers can use simulations of the planned research and then
proceed if the results are satisfactory [8].
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Quantitative criteria for accident tolerant fuel
Discussions from the previous section show that ATF needs to have certain properties
that make them tougher in accident conditions. Desired functions are that they should
withstand loss of active cooling for a much longer duration than existing fuels, with
the result that the existing safety margins will increase. This should help to enhance
performance of the fuel for a longer time. ATFs should be adaptable to existing
nuclear plants and allow for licensing for future reactors. They should also reduce
maintenance and operational costs. This section presents important quantitative
criteria for performance of ATFs in severe accident conditions. An important point is
that a number of cladding systems were discussed and these systems have their own
properties and features, and it may not be possible to obtain common quantitative
criteria for them [9]. However, efforts are taken to bring the best possible fit criteria.
Steam Oxidation temperatures: This parameter refers to the temperature at which
steam oxidation of the cladding begins. Fig 3.1 presents data on this temperature for
different cladding systems.
Figure 3.1 Steam Oxidation temperature [10]
As seen in the figure, the standard or conventional Zr alloy starts oxidation at a
temperature of 1273-1773 degree centigrade. In section 3 [3, 4] SiC has emerged as
an important source for cladding system. SiC system has a cladding oxidation
temperature of 1473-1873 degree centigrade. Hence, the first assessment is that the
oxidation temperature range for ATF of SiC should be in at a middle value of 1763
degree centigrade. However, one should also consider APMT system with Ti2AlC
system [5, 7]. It is seen that this system has an oxidation temperature resistance of
1323-1748 degree centigrade. The mean temperature of this system can be considered
as 1535 degree centigrade.
Oxidation time and film formation: These parameters indicate the exposure duration
when oxidation starts and the amount of oxidation is measured by the film formation
on the substrate. Materials with lower oxidation temperatures show weight gain due to
oxidation film formation. The objective is to prolong this duration to the maximum
value and delay film formation with weight gain. Fig 3.2 shows the results of samples
tested for weight gain when placed in a steam environment at 1200 degree centigrade
[11].
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Figure 3.2 Weight gain for samples at 1200 degree steam [11]
Samples that were tested are Si, Cr, Zr4 alloy, and SiO2. From the graph, it is seen
that Zr4 alloy starts to gain weight steadily from 0 seconds and this linear weight gain
continues for the full duration of the test that ran for 2000 seconds. Therefore, it is
clear that Zr4 alloy is perhaps not suitable for temperatures above 1200 degree
centigrade. The other alloys showed very little weight gain indicating that oxidation
does not occur and film does not precipitate for the full duration of 2000 seconds. As
seen in Fig 3.2, Si, Cr, and SiO2 showed marginal weight gain. Hence, the
quantitative data for ATF is that samples should have less than 5 mg/ dm2 when
exposed to a steam environment for 2000 seconds.
Young’s Modulus: The Young's Modulus is important in understanding the elastic
properties of materials. It gives the ratio of stress to strain and the value defines the
behavior under different load conditions. Materials with higher values are tensile and
they can resist deformation better. Young's modulus varies with temperature and at
higher temperatures, the material is more tensile and the value reduces. This aspect is
important in ATF since they operate at elevated temperatures. In a test, FeCrAl alloys
were prepared and tested. These alloys had compositions in weight % of Fe ranging
from 83.98 to 77.86, Cr from 12.99 to 16.06, and different compositions for Al, Mo.
Si, Nb, etc [12]. Fig 3.3 presents the graph of elastic modulus to temperature for these
metals.
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Figure 3.3 Young’s modulus of ATF [12]
From Fig 3.3, it is clear that Zr alloy shows a sharply decreasing Young’s modulus at
900 degrees centigrade, where the value is 45 Giga Pascals. APMT alloy samples
show the highest value of 140 GPa, while PM2000 sheet shows a value of 100 GPa.
Summary: The recommended specifications for ATF, based on [10, 11, 12] are as
follows.
Operating temperature range: 1535 degree centigrade
Weight gain from oxidation and film formation: 5 mg/ dm2 when exposed to a steam
environment for 2000 seconds.
Young' Modulus: 140 GPa at 900 degree centigrade.
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Summary
The paper examined the significance and need for ATF and discussed the current
status of research. The findings indicate that cladding materials made from SiC,
coatings of Cr, Si and Al3Ti alloys, composite matrix of Zr and ceramic layers,
micro-pellets of UO2-FeCrAl and FCM-FeCrAl systems, are under tests. However,
clear specifications of materials and specifications are not yet available. The paper
also presented quantitative criteria for important parameters such as minimum steam
oxidation temperature and weight gain from oxidation, and the Young's Modulus.
These specifications need rigorous testing before they are approved and their impact
on costs, compatibility, reliability, and other factors must be considers. The opinion
and feelings formed from the research is that the subject of ATF is very critical and
extensive research with live trails needs to be undertaken. Suggestions are to increase
funding and provide researchers with adequate facilities to take up urgent research
and development. Nuclear accidents in the past few years show that the cleanup costs
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and damages are very high. These costs and human suffering can be avoided with
ATF.
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