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1
Introduction
The Molten Salt Reactor Experiment (MSRE) at Oak Ridge
National Laboratory (ORNL) was a demonstration of a novel nuclear
reactor design. The reactor generated a power of ~ MW from nuclear
fission reactions in operations during 1965-1969. Graphite in the MSRE
reactor vessel acted as a moderator to slow down (or "moderate") the
high-energy neutrons released from fission events, to reduce the neutron
kinetic energies to low (or "thermal") values where the fission cross
section is much larger than it is at higher neutron kinetic energies. As in
any nuclear reactor, the combination of fissile filet and moderator in the
core generated a spatially distributed flow of neutrons, or a neutron flux,
which is the number density of neutrons passing through a unit area in a
unit time as a function of position in the reactor, time, and neutron
energy.
The following features made the MSRE design atypical. The
MSRE fuel was in a homogeneous molten fluoride salt medium, rather
than in solid nuclear idle! rod assemblies that are used in all commercial
nuclear power plants today. The filer was fissile uranium (first a 23su
charge, later 233U) and fissile plutonium (239Pu) contained as the uranium
tetrafluoride (UF4) and plutonium tetrafluoride (PuF4) salts in a molten
salt medium consisting predominantly of lithium, beryllium, and
zirconium fluorides (LiF, BeF2, and ZrF4), at an operating temperature of
approximately 650°C.
Another unusual feature was the lack of a separate coolant in the
core design. The reactor vessel contained only the molten salt filet and
the graphite moderator. Heat from the fission reactions elevated the
iTo sustain a critical chain reaction in an assembly of fissile material, each nuclear
fission event produces, for every thermalized neutron captured, approximately two
energetic neutrons (which, with moderation and loss mechanisms, result in approximately
one thermalized neutron to participate in a subsequent fission event). Energy is also
liberated, on the order of 200 million electron volts (MeV) per fission. The fissile nuclei
used in the MSRE were uranium-233 and -235 (233U, 35U) and plutonium-239 (239Pu).
10
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INTRODUCTION
11
temperature of the fuel salt and was removed by circulating the fuel salt
through a heat exchanger external to the reactor core. This design is to be
contrasted with that of commercial reactors, which circulate the coolant
(water or gas) in the core of solid fuel rods.
Figures I.l and I.2 show the layout of the MSRE. The fuel salts
were circulated in a loop containing the reactor vessel, a heat exchanger,
and a pump. The coolant salts circulated in a loop consisting of the heat
exchanger, a separate pump, and a radiator, from which the heat was
dissipated to the air (up a stack) by fans.
The reactor is housed in an isolated concrete building of ORNE
inside the Oak Ridge Reservation. The drain tanks and reactor vessel are
in a sealed containment system (Peretz, 1996c, p. I-6) surrounded by
concrete walls several feet thick to provide a primary radiation shield
between the radioactive materials and the environment (see Figure 1.31.
The building provides a secondary containment. The nearest location for
exposure to the public is the Bethel Valley Road outside ORAL (Peretz,
1996c, p. 1-36~.
The MSRE demonstrated the potential for a thermal neutron
breeder design. Introduction of thorium-232 (232Th) in the salt (as a
fluoride compound ThF4) would permit breeding (i.e., generation by
nuclear-induced transmutation) of 2 3U. Absorption of a low-energy (near
thermal) neutron by a 232Th nucleus transmutes it to 233Th, which
undergoes beta decay to protactinium-233 (233Pa), followed by a second
beta decay to produce 23 U. Breeding is achieved in a reactor when the
rate of production of fuel (233U) is greater than the rate of consumption of
fuel by nuclear fission inside the core. The ratio of the rate of production
and the rate of consumption defines the breeding ratio.
Although actual experiments with a thorium blanket were not
performed, experiments on the MSRE demonstrated the potential for a
next-generation design to breed 233U with a breeding ratio greater than
one. An on-line chemical processing system could have been installed (in
the fuel processing cell shown in Figure 1.2) to extract 233Pa from the salt
during operations to enhance the production of 233U. If the 233Pa were left
in the core, its absorption of thermal neutrons would reduce both the
neutron flux in the core and the breeding of 233U (Nero, 1979~. The
processing system that was installed was used successfully to recover
uranium by fluorination (Peretz, 1996c).
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12 AN EVALUATION OF DOE ALTERNATIVES FOR MSRE
~ REMOTE MAINTENANCE
REACTOR
CONTROL
ROOM
\
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CONTROL ROOM I
l
1. REACTOR VESSEL
2. HEAT EXCHANGER
3. FUEL PUMP
4. FREEZE FLANGE
5. THERMAL SHIELD
6. COOLANT PUMP
7. RADIATOR
8. COOLANT DRAIN TANK
9. FANS
10. FUEL DRAIN TANKS
11. FLUSH TANK
12. CONTAINMENT VESSEL
13. FREEZE VALVE
, -:...
FIGURE 1.1 Arrangement of the principal components of the MSRE. SOURCE:
Modified from Peretz (1996c, Figure 1.1~.
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INTRODUCTION
15
A more comprehensive discussion of the reactor can be found in
many sources (see, for example, Benedict et al., 1981, p. 10; Knief, 1992,
pp.318-321; Weinberg, 1 994, pp. ~ 25-127; Peretz, ~ 996c, and references
therein).
CURRENT STATUS OF THE MSRE
After shutdown in 1969, the molten fluoride fuel salt and a batch
of flush salt (used to "flush" the system) were drained into three drain
tanks, two fuel salt tanks and one flush salt tank, and were permitted to
solidify. The tanks are located in a hermetically sealed (welded shut)
enclosure containing both the drain tank cell and the reactor cell (see
Figures I.~-~.4~. The major chemical constituents of the salts are lithium,
beryllium, and zirconium fluorides; the radioactive species are actini`des,
their decay products, and fission products.
Radiation emitted from the decay of radioactive species within
the salt has interacted with the salt medium to cause radiolytic formation
of both F2 (fluorine) and UFO (uranium hexafluoride) gas. These gases
have diffused out of the solid salt. This diffusion may have been aided by
periodic annealing operations at elevated temperature that were
performed in the 1970s and 1980s. Over time, headspace gases in the
tanks have migrated into the reactor vent piping and the off-gas vent trap
system to the interior of the activated charcoal bed (ACB; see Figure ~ .5)
through valve openings. In recent years, the migration of UFO gas through
the system apparently has resulted in the deposition of uranium fluoride
solids in the ACE, and such deposits may be present on the interior walls
of piping and in the drain tanks. In October 1996, the vent piping was at
an overpressure of approximately one atmosphere, due to F2 and UFO
gases, with solid plugs of uranium fluoride deposits inferred from
pressure differences in piping runs leading from the drain tanks to the
ACE at the end of the vent line.
The facility's two immediate hazards gas buildup and solid
uranium fluoride deposits are being addressed by work in progress at
ORNL. A "reactive gas removal" action has been designed to unblock or
bypass the plugged piping system and collect the chemically reactive F2
and UFO gases in traps. Pumping initiated in November 1996 has relieved
the overpressure in some piping, although nonvolatile plugs have prevented
~ ~ ~ V ~ A ~
16 AN EVALUATION OF DOE ALTERNATIVES FOR MSRE
STEAM OUTLET-
GONDENSATE RETURN -
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HANGER GABLE
_~` INSTRUMENT THIMBLE
~ FUEL SALT DRAIN TANK
,.
~ TANK FILL LINE
11 ~
—THIMBLE POSITIONING RINGS
FIGURE 1.4 Sketch of an MSRE fuel drain tank, made of a nickel alloy
(Hastelloy N), showing bayonet thimbles attached to a headspace drome. The
flush drain tank lacks the bayonet thimbles. SOURCE: Peretz (1996c, Figure 1.3~.
17
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18
AN EVALUATION OF DOE ALTERNATIVES FOR MSRE
complete access to all system piping, including that closest to the drain
tanks. A robotic vacuuming operation is being designed to extract the
uranium deposits and charcoal from the top of the ACB trap.
The uranium fluoride salts in the three drain tanks are a more
long-term concern. These fluoride salts are not in a favorable
configuration for long-term storage. The salts are unstable due to
radiolysis that continues to occur, liberating reactive gases. The fissile
materials in the tanks could form a critical configuration if a moderator
such as water entered the system. The probability of this happening in the
long term must be considered if the salts are to be stored indefinitely in
the present location (see Figures 1. I- ~ .3~. In particular, the drain tank cell
is capable of being flooded if the hermetic seal is lost because the natural
water table is above the tanks (Peretz, 1996c). Waterflooding could occur
in the absence of active ground water pumping operations, which are
currently practiced. Because a nuclear criticality excursion cannot be
ruled out absolutely under these conditions, the salts in the present mode
of tank storage pose a long-term hazard, even though it is small.
The Department of Energy (DOE), following the Comprehensive
Environmental Response, Compensation, and Liability Act (CERCLA)
regulatory process, is exploring technically feasible operations designed
to reduce the hazard posed by the salts in these tanks. A variety of
technical alternatives have been identified to date (Peretz, 1996c) for salt
removal, separation of uranium, and ultimate disposition of all waste
products.
Seven primary alternatives, some with subalternatives, were
identified by Peretz (1996c). They are summarized as follows: (1) do
nothing, (2) enhance the present storage, (3) treat the salts as spent
nuclear fuel awaiting eventual storage in a federal repository, (4) treat the
salts as transuranic waste bound for the Waste Isolation Pilot Plant, (5)
electrorefine the salts, (6) reuse the salts elsewhere in the DOE complex,
and (7) use interim storage at a DOE site. Alternatives 4 through 7
include a possible fluorination step to remove uranium. Chapter 7 lists
these alternatives in more detail.
These technical alternatives are also considered in the larger
context of a waste management strategy with a view to ultimate disposal
options, such as geologic storage in a repository. The strategic conclusion
of Peretz (1996c) is that on-site interim storage is the preferred strategy
for the foreseeable future, provided the materials in question cannot be
INTRODUCTION
19
reused beneficially elsewhere in the DOE complex. The particular
technical alternative recommended by Peretz (1996c) is a melting and
fluorination process to strip uranium from the salt as UFO gas. The
uranium fluoride would be converted to an oxide and stored with other
233U inventories, while the stripped salt would be stored in a "gettered"
mode in shielded cells.
This alternative, and others, would involve remediation work using
the MSRE facility and equipment. Project personnel have begun testing and
maintenance, such as replacement of deficient pressure sensors. However,
the operability and status of some components are unknown and
contributes to the panel's recommendation for a stepwise remediation
approach. Present DOE practices require an assessment of the integrity of
all components of the MSRE that would be vital to a remediation operation.
These considerations are at a stage where relevant scientific and
technical information now under development on the identified
alternatives will provide input to a future binding regulatory decision on
the particular remediation option selected for the MSRE salts. This
decision involves selection of the remediation alternative considered to
be the best, as judged against all CERCLA criteria.2
ROLE OF THE NATIONAL RESEARCH COUNCIL
At the request of DOE, the National Research Council (NRC),
under the auspices of the Board on Radioactive Waste Management
(BRWM), has undertaken a study to provide an independent technical
review of the alternatives offered in Peretz (1996c~specifically, to
examine the process by which these alternatives for MSRE salt treatment
and disposition were identified and used in decision making. The
complete Statement of Task describing the scope of this study is
reproduced in Box 1.1. The study was undertaken by the Molten Salt
Pane} (hereafter referred to as the panel) of the Committee on
Remediation of Buried and Tank Wastes (CRBTW). The present report,
written by the panel, represents consensus positions reached as a result of
2The CERCLA criteria are the following: overall protection of health and environment,
compliance with applicable regulations, long-term effectiveness, reduction of hazard,
short-term effectiveness, implementability, and cost.
20
AN EVALUATION OF DOE ALTERNATIVES FOR MSRE
deliberations and discussions. Two meetings (in September and October
1996),3 parts of which were open to the public, were held to solicit input
from and interaction with DOE and its contractors. The pane} also heard
from current and retired Oak Ridge employees, a State of Tennessee
representative, and interested members of the public.
The panel used many available background materials. The
problems and proposed solutions associated with the salt tanks have been
discussed in some detail at various technical meetings (Peretz, [996a,b),
in the reports of a Senior Review Board review (February 20, 1995), and
in briefings on MSRE fuel salt disposition to the CRBTW during 1996.
Chemical and technical aspects of the problem, including the current
state of the molten salt reactor fuel and flush salts, tanks, and related
equipment (e.g., associated tubing, valves, and sensors), have been
discussed and summarized in Peretz (1996c) and references therein.
Presentations to the pane} (Rushton et al., 1996a,b) have reviewed the
history of facility operations and have outlined a targeted timeline for
safe removal and disposition of the MSRE fuel and flush salts that is
consistent with CERCLA requirements. The future schedule is intended
to encompass actions subject to the approval of the two agencies with
regulatory authority over these actions, the State of Tennessee and the
U.S. Environmental Protection Agency.
SCOPE AND ORGANIZATION OF THIS REPORT
To address the three key questions posed in the Statement of
Task (Box ~.1), this report is organized as follows.
Some of the scientific challenges common to most of the
alternatives are discussed in this report first. Radiolysis and nuclear
reactions are treated in Chapter 2. Fluoride salt chemistry, partitioning,
and system corrosion are treated in Chapter 3. The assessment of the
present condition of the salts and the drain tanks that is contained in these
two chapters is relevant to any remediation approach.
Other challenges are discussed as they would pertain to the
execution of a remediation plan. In Chapter 4, the logical process steps
are presented, from which the panel develops a preferred, stepwise
3The September meeting included a tour of the MSRE facility.
INTRODUCTION
21
22
AN EVALUATION OF DOE ALTERNATIVES FOR MSRE
approach based on the understanding that some decisions are best
postponed until after further information on the system is acquired.
Chapter 5 presents selected technical processing approaches to strip
uranium, with commentary on the relative advantages and disadvantages
of each. Issues pertaining to nuclear criticality safety provide one of the
most significant constraints in some of the future processing options;
therefore, these considerations are explicitly treated in Chapter 6.
The scientific and technical background information in Chapters
2 through 6 is important to the assessment of strategic alternatives, which
are discussed in Chapter 7. Several of the various alternatives in Peretz
(1996c) are compared and contrasted to provide the basis for a
recommended alternative.
Chapter ~ addresses the management of potential hazards
associated with MSRE salt cleanup operations. This chapter also lists
several specific information-gathering activities recommended to reduce
hazards and provide experimental evidence to better inform the decision-
making process.
The final chapter (Chapter 9) contains a summary discussion
with commentary on the waste management strategy appropriate for the
molten salt cleanup efforts. To the extent that future experimental
evidence confirms the anticipated condition of the salt and drain tanks, a
preferred approach is recommended for consideration. The three
questions posed in the Statement of Task are also answered explicitly.
