Reducing Actinide Production Using Inert Matrix Fuels

42 Pages·2017·6.39 MB·English

by  Mark Deinert

Project No. 12-3595
Reducing Actinide Production
Using Inert Matrix Fuels
Mission Supporting Transformative Research
Mark Deinert
Colorado School of Mines
Dan Vega, Federal POC
Jon Carmack, Technical POC
NEUP – Project: 12-3595
Title: Reducing Actinide Production Using Inert Matrix Fuels
University Name: The Colorado School of Mines
Principle Investigator: Dr. Mark Deinert
Reporting Period: Final Report

Abstract: The environmental and geopolitical problems that surround nuclear power stem
largely from the longlived transuranic isotopes of Am, Cm, Np and Pu that are contained in spent
nuclear fuel. New methods for transmuting these elements into more benign forms are needed.
Current research efforts focus largely on the development of fast burner reactors, because it has
been shown that they could dramatically reduce the accumulation of transuranics. However,
despite five decades of effort, fast reactors have yet to achieve industrial viability. A critical
limitation to this, and other such strategies, is that they require a type of spent fuel reprocessing
that can efficiently separate all of the transuranics from the fission products with which they are
mixed. Unfortunately, the technology for doing this on an industrial scale is still in development.
In this project, we explore a strategy for transmutation that can be deployed using existing,
current generation reactors and reprocessing systems. We show that use of an inert matrix fuel to
recycle transuranics in a conventional pressurized water reactor could reduce overall production
of these materials by an amount that is similar to what is achievable using proposed fast reactor
cycles. Furthermore, we show that these transuranic reductions can be achieved even if the
fission products are carried into the inert matrix fuel along with the transuranics, bypassing the
critical separations hurdle described above. The implications of these findings are significant,
because they imply that inert matrix fuel could be made directly from the material streams
produced by the commercially available PUREX process. Zirconium dioxide would be an ideal
choice of inert matrix in this context because it is known to form a stable solid solution with both
fission products and transuranics.

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Table of Contents

Page 1: Introduction
Page 6: Project aims
Page 7: Methods
Page 15: Results and discussion
Page 22: Conclusions
Page 23: Supplementary information
Page 34: References
Page 38: Appendix 1 – List of Publications

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List of Figures

Page 1: Figure 1. How to reduce transuranic inventories.
Page 2: Figure 2. Proposed recycle strategy.
Page 5: Figure 3. The inert matrix fuel cycle.
Page 6: Figure 4. Fuel temperatures.
Page 9: Figure 5. Fuel assembly configuration.
Page 10: Figure 6. Axial distribution of burnable absorber in fuel.
Page 12: Figure 7. Effect of burnable absorber.
Page 14: Figure 8. Moderator Thermal Reactivity Coefficients.
Page 15: Figure 9. Void Reactivity Coefficients.
Page 16: Figure 10. Axial power profile.
Page 17: Figure 11. Axial temperature profile.
Page 18: Figure 12. Radial core power profile.
Page 19: Figure 13. Assembly level power.
Page 20: Figure 14. Reactor Criticality.
Page 21: Figure 15. Time to safe storage under loss coolant conditions in a spent fuel pool.
Page 23: Figure A. Reactor Core configuration.
Page 24: Figure B. Fuel transuranic content as a function of time.
Page 25: Figure C. Plutonium content as a function of residence time.
Page 26: Figure D. Heavy metal content in the inert matrix fuel cycle.
Page 28: Figure E. Thermal conductivity coefficients for reactor materials.
Page 30: Figure E. Fuel reactivity coefficient at beginning of cycle.
Page 31: Figure F. Fuel reactivity coefficient at end of cycle.
Page 32: Figure G. Moderator Thermal Reactivity Coefficients.
Page 33: Figure H. Void Reactivity Coefficients.
Page 34: Figure I. Assembly level power.

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Final report

NEUP – Project: 12-3595
Title: Reducing Actinide Production Using Inert Matrix Fuels
University Name: The Colorado School of Mines
Principle Investigator: Dr. Mark Deinert
Reporting Period: Final Report

Introduction. The environmental and geopolitical problems that surround nuclear power stem
largely from the long-lived transuranic isotopes of Am, Cm, Np and Pu that are contained in
spent nuclear fuel [1, 2]. If these four elements could be eliminated, many of the issues
associated with nuclear power would disappear. At present, the only commercially viable option
for reducing the buildup of any of these elements is to mix recycled plutonium with uranium to
produce a ‘mixed oxide’ fuel. However, due to neutron capture in the uranium, these fuels also
produce transuranics while in the reactor, and are therefore only marginally effective at limiting
overall transuranic production. Current research efforts
focus largely on the development of fast burner reactors as
the optimal system for transmutation [2, 3], and Fig. 1
illustrates why; they can dramatically reduce the
accumulation of transuranics. However, despite five
decades of effort, fast reactors have yet to achieve industrial
viability. In addition, transmutation in fast reactors requires
a type of spent fuel reprocessing that can efficiently
Figure 1. How to reduce transuranic
separate all of the transuranics from the fission products inventories. The cumulative discharge is
shown for three fuel cycles over 100 years
with which they are mixed. Unfortunately, the technology (t=0 corresponds to 2012). The UOX cycle
is what is currently used in the US. The
for doing this on an industrial scale is still in development inert matrix (IMF) and fast burner cycles
assume that transuranics from spent UOX
[4, 5]. As a result, there is a pressing need for would be recycled either as single pass IMF
or continuously in fast reactors.
.
transmutation strategies that can be deployed with existing technologies.

Another approach would be to recycle the transuranics in a conventional reactor but using an
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‘inert matrix’ fuel which is free of uranium. This fuel would allow for transuranic consumption
without any production. Because of its properties, zirconium dioxide has received considerable
attention for this application [6-16]. Results have shown that using an inert matrix fuel to
recycle transuranics in a conventional pressurized water reactor could reduce overall production
of these materials by an amount that is similar to what is achievable using proposed fast reactor
cycles, Fig. 1 [17, 18].

While the inert matrix results shown in Fig. 1 are compelling, they also assume that transuranics
and fission products can be separated from one another. However, data (discussed below) shows
that the transuranic reductions shown in Fig. 1 can be achieved even if the fission products are
carried into the inert matrix fuel along with the transuranics. This is important because the
commercially available PUREX process produces streams of plutonium as well as a mixture of
fission products and the transuranics Am, Cm, Np. These streams can be blended with zirconium
dioxide to make inert matrix reactor fuel, Fig. 2. Past work [11, 19] has shown that zirconium
dioxide will form a stable solid solution with
fission products and transuranics. This is
significant because it opens the door to a
transmutation strategy that uses currently
available technologies to reduce production
Pu
x% IMF
PUREX
of Pu, Am, Cm, and Np. The goal of the y% UOX
FP, Am

Cm, Np
proposed work is to demonstrate that a
reactor running this type of recycle strategy
could be licensed for commercial operation.
The basic requirements for licensing were
Figure 2. Proposed recycle strategy. Here FP
laid out in the certification study that was
stands for non-gaseous/non-volatile fission products.
The optimal values for x, y and uranium fuel
undertaken for the use of mixed oxide fuel in
enrichment will be determined as part of the study.
The streams of Pu and FP + (Am, Cm, Np) can be
the US [20]. In the context of the present
blended in different ratios to give inert matrix fuel with
different reactivities.
study we will perform a parametric analysis
to establish the space of reactor configurations, and fuel compositions, that would allow a reactor
to recycle its own transuranics under the following necessary constraints: the fuel remains within
allowable thermal limits, the reactivity coefficients for the reactor remain negative, and the
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reactor remains in an under-moderated state during operation.

Importance. There are two standard approaches by which to limit the production of transuranics
in nuclear power: reduce the use of uranium in a nuclear reactor or recycle and transmute the
transuranics elements that are contained in spent fuel. Many strategies have been proposed over
the decades that exploit one or both of these approaches. However, and after decades of
research, all of these approaches require reactor systems or separations technologies that have yet
to achieve commercial viability [21-23].

The thorium cycle is an example of how to limit the use of uranium in a core, and thereby reduce
the production of transuranic elements per unit of electricity generated. Here fissile 233U is bred
by neutron capture in natural thorium. Reprocessing is then used to remove the 233U and produce
fresh fuel. While simulations have demonstrated that such a fuel cycle could work, and thorium
fueled reactors have operated, the reprocessing technology needed for this approach has never
been demonstrated at the industrial scale. Another problem with this approach is the
proliferation risk associated with separated quantities of 233U. In practice 238U would need to be
added to the thorium fuel to dilute the 233U below 19.6 a/o required by proliferation standards
[24, 25]. However, the presence of significant amounts of 238U in the fuel reduces the ability of
this approach to significantly impact the overall transuranic production. This would also hold if
thorium was used as a fertile free matrix in which transuranics from conventional spent uranium
dioxide fuel could be recycled.

It is well established that the current option of recycling plutonium using mixed oxide fuels as
practiced in France has only a marginal impact on the overall production of transuranics [2].
Variations on this approach, such as CORAIL, could have a significant effect by using a multiple
recycle approach that uses all of the transuranic elements either directly in the fuel assemblies or
as targets in a thermal spectrum core [26]. The CORAIL approach is particularly attractive
because past work has shown that continuous recycle of all high actinides using this approach
would come close to achieving steady state transuranic inventories. In addition, the operational
performance of the mixed oxide component of the CORAIL assemblies is well understood for
plutonium recycle [26] and other work suggests that it would be similar if all the transuranics
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were recycled. The problem here is that CORAIL requires separation of plutonium and minor
actinides from the fission products with which they are entrained. As of yet, no process exists
that has been demonstrated to be commercially viable for this application. If this is not done, the
fission products would also need to be recycled and would build to a point where they would
produce a mixed oxide fuel that would poison the reactor.

Fast reactor systems have perhaps received the most attention as a means for limiting the
production of transuranics. These reactors are attractive not only for their ability to cap
transuranic inventories, Fig. 1, but also because they can be used to breed fissile material and
reduce the need for uranium mining. As a result, fast reactors have been the focus of serious
research and design efforts in the US, UK, France, Russia, Japan, India and more recently China.
The International Atomic Energy Agency lists 19 fast reactors as having been built and operated
to date [27]. Of these, seven have suffered from sodium leaks, caught fire or incurred partial
core meltdowns. The most successful commercial fast reactor to date has been the Russian
BN600 which has operated since 1980 and achieved a lifetime capacity factor of 77%. However,
even this reactor has had two fires in its sodium cooling system [27-29]. At this time, no fast
reactor design is licensed for commercial operation in the west and none has demonstrated
performance that would be commercially viable when compared to conventional light-water
technology. While there is little question within the nuclear engineering community that fast
reactors will eventually be commercialized, fast reactor cycles also require efficient separation of
transuranics from the fission products with which they are entrained. Here again, the technology
is not yet deployable at industrial scale.

Work on thorium, CORAIL and fast reactor systems has shown that it is essential for all of the
transuranic elements to be recycled if serious reductions in nuclear power’s long term radioactive
signature are to be achieved. Unfortunately this requires a method by which the transuranic
elements can be separated from the fission products. While the PUREX process can do this for
plutonium [30], no industrial scale process currently exists that can separate the remaining Am,
Cm, and Np from the fission products with the required efficiency [4, 5]. As a result, the
approaches described above (as well as others not mentioned) all rely on technologies that
remain immature. This situation poses a problem for the US commercial nuclear industry
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because it means that interim storage is the only near-term option for spent nuclear fuel and this
view has been advanced in influential reports out of MIT and Harvard [22, 23, 31].

Inert matrix fuels provide an alternative approach that could be deployed using both current
generation light-water reactors and PUREX reprocessing. If commercially viable, this presents a
significant advance over other transmutation approaches that rely on future technologies. The
idea here is that PUREX would be used to create three material streams: uranium, plutonium,
and a mixture of higher actinides + non-gaseous/non-volatile fission products. The latter two
would then be recombined and blended with zirconium dioxide to create inert matrix fuel, Fig. 3.

Figure 3. The inert matrix fuel cycle. Spent fuel from a conventional light-water
reactor would be reprocessed and the transuranics Am, Cm, Np and Pu stripped, and
blended with a uranium free matrix. The resulting fuel would be placed back into the
light-water reactor. Because it contains no uranium, the inert matrix fuel form allows
for the consumption of transuranic waste without any additional production. The
inert matrix fuel would also reduce the overall amount of uranium in the core. The
percentage of the core that is uranium dioxide (UOX) or inert matrix fuel (IMF) is a
design parameter.

Work by our group has shown that inert matrix fuels, loaded with transuranics alone, would
exceed their allowable thermal limits at beginning of life, Fig. 4 (Left). However, we have also
been able to show that varying the axial composition of the fuel can be used to flatten the power
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profiles so that the fuel temperatures remain within licensable limits, Fig. 4 (Right), and that the
core power profile has a peak to average ratio ~1.3. Importantly, full core simulations also
showed that a pressurized water reactor running this type of fuel cycle would maintain negative
void and thermal reactivity coefficients, which are also required for licensing [32, 33].

2500 1500 Figure 4. Fuel temperatures.
1400 Left: centerline fuel temperatures
before axial grading with burnable
elvin]2000 elvin]11230000 poisons. Right: After grading.
K K The results are for an inert matrix
e [ e [1100
atur1500 atur1000 fuel with transuranics only. Full
er er core simulations of a mixed core
mp mp 900 PWR were performed using
Te1000 Uranium Oxide Te 800 Uranium Oxide MCNPX 2.7. Axial grading kept
Inert Matrix 700 Inert Matrix fuels within allowable thermal

500 600 limits.
0 100 200 300 400 0 100 200 300 400
Axial distance [cm] Axial distance [cm]
If
the non-gaseous/non-volatile fission products are also recycled, they will act as a burnable
poison in their own right and depress the peaking seen in Fig. 4 (Left). By varying the ratio of
Pu / (higher actinides + non-gaseous/non-volatile), the axial power profile of the fuel pins could
also be adjusted without the need for additional poison. Even so, the fission products constitute a
considerable neutron sink and care must be taken to ensure that a reactor running this type of fuel
cycle would in fact remain critical between refuelings.

Project Aims. The question that we are addressing is whether a reactor with a mixed core of the
type described would be licensable for commercial operation. The issues that need to be
addressed here were all laid out in the certification study that was undertaken for mixed oxide
fuel in the US [20]. In simple terms this analysis requires: i) A neutronic analysis to determine
whether the reactor will maintain negative reactivity coefficients as well as allowable power
distributions throughout its operational cycles; ii) A thermal-hydraulic analysis to show that the
fuels will remain within allowable thermal limits; iii) A simulation of fuel decay heat (as a
function of fuel burnup) as this would affect both in-pool storage and accident performance; iv)
An accident analysis. The accident analysis is unnecessary unless the results of the first three
studies show that the performance of the inert matrix fuels, and the cores in which they are run,
are within allowable limits. This project has focus on these analyses.

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