PB98-964002
EPA 541-R98-018
September 1998
EPA Superfund
Record of Decision:
Oak Ridge Reservation (USDOE)
Molten Salt Reactor Experiment
(MSRE) Facility
Oak Ridge, TN
7/7/1998
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DOE/OR/02-1671&D2
Record of Decision for Interim Action
to Remove Fuel and Flush Salts
from the Molten Salt Reactor Experiment Facility
at the Oak Ridge National Laboratory,
Oak Ridge, Tennessee
-------
DOE/OR/02-1671&D2
Record of Decision for Interim Action
to Remove Fuel and Flush Salts
from the Molten Salt Reactor Experiment Facility
at the Oak Ridge National Laboratory,
Oak Ridge, Tennessee
Date Issued—June 1998
Prepared by
Jacobs EM Team
125 Broadway Avenue
Oak Ridge, Tennessee
Prepared for the
U.S. Department of Energy
Office of Environmental Management
Environmental Management Activities at the
Oak Ridge National Laboratory
Oak Ridge, Tennessee 37831
managed by
Bechtel Jacobs Company LLC
for the
U.S. Department of Energy
under contract DE-AC05-98OR22700
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PREFACE
This Record of Decision for Interim Action to Remove Fuel and Flush
Salts from the Molten Salt Reactef Experiment Facility at the Oak Ridge
National Laboratory, OakRidge, Tennessee (DQE/QR/Q2-1611&.D2) was
prepared in accordance with requirements under the Comprehensive
Environmental Response, Compensation, and Liability Act of 1980. The
U.S. Department of Energy, U.S. Environmental Protection Agency, and
the state of Tennessee agree here to select the action for removing fuel
and flush salts and placing the salt in a more controlled storage condition
until final disposition of the salt is arranged. Work on this task was
performed under Work Breakdown Structure 1.4.12.6.2.01 (Activity
Data Sheet 3700, "Molten Salt Reactor Experiment D&D Support").
This document presents a description of the selected remedy, which
includes removing flush salt and fuel salt from their respective storage
containers in the Molten Salt Reactor Experiment facility, removing
uranium from the salts, treating the uranium to form an oxide for safer
storage, placing the uranium oxide into storage, containerizing the fuel
and flush salts without uranium, and temporarily storing this salt at the
Oak Ridge National Laboratory until final disposition of the salt. This
document relies on and is consistent with information in the Feasibility
Study for Fuel and Flush Salt Removal from the Molten Salt Reactor
Experiment at the Oak Ridge National Laboratory, Oak Ridge, Tennessee
(DOE/OR/02-1559&D2), the Interim Action Proposed Plan for Fuel and
Flush Salt Disposition from the Molten Salt Reactor Experiment, Oak
Ridge National Laboratory, Oak Ridge, Tennessee (DOE/OR/02-
1601&D3), and Evaluation of the U.S. Department of Energy's
Alternatives for the Removal and Disposition of Molten Salt Reactor
Experiment Fluoride Salts prepared by the National Research Council in
1997.
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ACRONYMS AND ABBREVIATIONS
ARAR
Be
CERCLA
Ci
D&D
DOE
EPA
FFA
FS
ft
g
HF
kg
km
Ib
Li
m
MSRE
NEPA
ORNL
ORR
ppm
ROD
TDEC
TRU
U
UF«
WIPP
Zr
applicable or relevant and appropriate requirement
beryllium
Comprehensive Environmental Response, Compensation, and Liability
Act of 1980
curie
decontamination and decommissioning
U.S. Department of Energy
U.S. Environmental Protection Agency
Federal Facility Agreement
feasibility study
foot
gram
hydrogen fluoride
kilogram
kilometer
pound
lithium
meter
Molten Salt Reactor Experiment
National Environmental Policy Act of 1969
Oak Ridge National Laboratory
Oak Ridge Reservation
parts per million
record of decision
Tennessee Department of Environment and Conservation
transuranic
uranium
uranium tetrafluoride
Waste Isolation Pilot Plant
zirconium
JT00869709.IBH/CJE
111
June 3. 1998
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PART 1. DECLARATION
JT00869709.1BH/CJE June 3. 1998
preceeding blank page omitted
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SITE NAME AND LOCATION
U.S. Department of Energy
Oak Ridge Reservation
Molten Salt Reactor Experiment Facility—Building 7503
Molten Salt Reactor Experiment Decontamination and Decommissioning Support
Oak Ridge, Tennessee
STATEMENT OF BASIS AND PURPOSE
This record of decision (ROD) presents the selected interim remedial action for addressing
fuel and flush fluoride salts from three drain tanks formerly used as part of the Molten Salt
Reactor Experiment (MSRE). The tanks are located in the MSRE facility (Building 7503) at the
Oak Ridge National Laboratory (ORNL) on the U.S. Department of Energy (DOE) Oak Ridge
Reservation (ORR). Remediating the MSRE facility is a high priority because of the
unacceptable risk associated with the highly radioactive salt stored in the drain tanks. The
location, condition, and age of the equipment connected to the tanks and the chemistry of the salt
make control of safety factors difficult. The objective of this interim action is to reduce potential
on- and off-site risk from the salt.
This interim action was chosen in accordance with the Comprehensive Environmental
Response, Compensation, and Liability Act of 1980 (CERCLA), as amended by the Superfund
Amendments and Reauthorization Act of 1986 (42 United States Code, Sect. 9601 et seq.) and,
to the extent practicable, the National Oil and Hazardous Substances Pollution Contingency Plan
(40 Code of Federal Regulations 300). The ROD is based on the Administrative Record for this
site.
DOE issues this document as the lead agency. The U.S. Environmental Protection
Agency (EPA) and the Tennessee Department of Environment and Conservation (TDEC) are
support agencies as parties to the Federal Facility Agreement (FFA) for this response action.
DOE and EPA have jointly selected the remedy for the MSRE fuel and flush salts removal.
TDEC concurs with the selected remedy.
JT00869TO9.IBH/CJE 1-3 June 3. 1998
preceeding blank page omitted
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ASSESSMENT OF THE STUDY AREA/OPERABLE UNIT
A streamlined risk assessment was conducted to determine whether current or future
remedial actions are necessary to protect human health and the environment if current institutional
controls are removed. The scenarios considered include on- and off-ske receptors. The risk
assessment demonstrates that without institutional controls the salts in the MSRE drain tanks pose
an unacceptable risk to human health and the environment now and in the future. Thus a
response action is required to address the salt stored in the three drain tanks at the MSRE facility.
The objective of this interim action is to reduce current potential on- and off-site risk from the
salts, pending final action.
Actual or threatened releases of hazardous substances from the MSRE facility that are not
addressed by implementing the response action selected in this ROD may present an unacceptable
risk to public health, welfare, and the environment.
DESCRIPTION OF SELECTED REMEDY
The selected interim remedial action includes melting and chemically treating the salt in
the drain tank cell, separating the uranium from the salts, transferring the uranium to the ^U
repository at ORNL, packaging the residual salt, and placing the salt in interim storage at ORNL
until arrangements are made for final disposition. Specific details and methods for this interim
remedial action will be included in the remedial design and remedial action plans. As the salt
melts in a drain tank, the molten salt will be treated with hydrogen fluoride (HF) to balance salt
chemistry. The uranium in the salts will then be removed from the salt and converted to an oxide
that is chemically stable and compatible with long-term storage at the B3U repository at ORNL
Building 3019 and managed as a part of the existing ^'U repository inventory. The residual salt
will be stabilized/packaged to control fluorine gas generation and the containers placed in interim
storage. The location of interim storage will be at an existing storage facility at ORNL.
Placement of the salt for its final disposition will be documented in a subsequent filial CERCLA
decision document and, as appisprMe, a National Environmental Policy Act of 1969 (NEPA)
decision document. These future decisions will incorporate full public participation and will be
based on the existing feasibility study (FS).
JT00869709.IBH/CJE 1-4 June 3, 1998
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After removal of salts from the MSRE drain tanks, the tanks and associated equipment
will be managed in place as part of the facility maintenance program. The storage tanks and
reactor components will be addressed as part of a subsequent decontamination and
decommissioning (D&D) action of the building.
STATUTORY DETERMINATIONS
This interim action protects human health and the environment, complies with federal and
state requirements that are legally applicable or relevant and appropriate requirements (ARARs),
and is cost-effective. Within its limited scope, this interim action uses permanent solutions and
alternative treatment technologies to the maximum extent practicable by removing the salts from
the MSRE drain tanks, treating the salts to remove the uranium, and stabilizing/packaging the
salts for final disposition. Therefore, the selected Interim remedy satisfies the statutory
preference for remedies employing treatments that reduce toxicity, mobility, or volume as a
principal element. Disposal and, if necessary, further treatment of MSRE salts after the uranium
has been removed will be performed as part of another action. This interim action addresses the
principal threat from criticality or release of contaminants into the environment posed by the salts.
stored in the MSRE drain tanks. Removal of radioactive salts will permit the remaining
structures to be included in a later action. Because this is an interim action ROD, review of this
facility will continue as DOE develops final remedial alternatives for D&D of Building 7503.
nXX>S69709.!BH/aE 1-5 June 3. 1998
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APPROVALS
Rodney-R. Nelson, Assistant Manager
U.S., Department of Energy
Oak Ridge Operations
Date
Earl C. Leming, Director
U.S. Department of Energy Oversight Division
Tennessee Department of Environment and Conservation
Date
Richard D. Green, Director
Waste Management Division
U.S. Environmental Protection Agency—Region 4
Date
JT00869709.IBH/CJE
1-6
June 3, 1998
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PART 2. DECISION SUMMARY
JTOO«69709JBH/CJE June 3, 1998
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SITE NAME AND LOCATION
The MSRE site is located in Roane County, Tennessee, on the DOE ORR approximately
1 km (0.6 miles) south of the ORNL main plant across Haw Ridge in Melton Valley. The ORNL
main plant is approximately 24 km (15 miles) west of Knoxville, Tennessee, and 16 km
(10 miles) southwest of the Oak Ridge, Tennessee, business center (Fig. 2.1).
The MSRE reactor and associated components are located in cells beneath the floor in the
high-bay area of Building 7503. The MSRE site with Building 7503 and other support buildings
are located at the intersection of Melton Valley Road and High Flux Isotope Reactor Access Road
(Fig. 2.2).
SITE DESCRIPTION, HISTORY, AND ENFORCEMENT
ACTIVITIES
Building 7503 was constructed in 1951 to contain the Aircraft Reactor Experiment and
expanded in 1955 for the Aircraft Reactor Test, which was canceled in September 1957. In
1961, experimentation on a molten salt reactor was revived at MSRE to develop a commercial
molten salt breeder reactor. Adjacent buildings supported the MSRE operation. The reactor,
using ^'U as fuel, achieved criticality on June 1, 1965. In August 1968, the ^U fuel was
replaced with B3U. The reactor operation permanently shut down December 12, 1969.
The MSRE reactor loop consisted of a reactor vessel, primary heat exchanger, pump,
associated piping, and an off-gas system (Fig. 2.3). During operation, the fluoride salt mixture
containing uranium fuel was heated to a liquid state. The molten salt was transferred from the
fuel drain tanks into the reactor circuit and criticality would occur in the reactor vessel. Fuel
salt, further heated by the nuclear reaction, exited the reactor vessel to the heat exchanger to
transfer excess heat to a secondary fluoride coolant salt. When the reactor was shut down, fuel
salt was removed from the reactor circuit by allowing it to drain by gravity back into the fuel
drain tanks. To remove residual fuel salt from the reactor circuit, molten flush salt was
circulated through the reactor circuit and returned to the flush salt drain tank. At the time
operations ceased, the fuel and flush salts were allowed to cool and solidify in the drain tanks.
The fluoride salt used for the fuel and flush salts in MSRE is generally similar except for
the uranium fuel and other radionuclide content differences. After shutdown, the fluoride fuel
salt and possibly the flush salt released fluorine and uranium hexafluoride gases into the drain
JT00867709 1BH/CJE 2-3 '"•* 3. '998
preceeding blank page omitted
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LEGEND
I : •-;. | DOE Oili Rrtloe FHnrvillon
^^| Prineipil OR ft
V&Sa Populition Centers
\
MODIFIED FROM DOE 1993
Fig. 2.1
DOE Oak Ridge Reservation and vicinity
DOE • ORNL. Molten Silt Reactor Experimtnl • 0>k Ridge. Tennessee
DOCUMENT ID 35H830
OOB6-40 / MSRE
OBAWINO ID:
• 7.U238DWG
DRAWING DATE.
FEBRUARY to, tggs SB
-------
Fig. 2.2
Aerial photograph of the MSRE site
DOE - ORNL. Molten Salt Reactor Experiment - Oak Ridge. Tennessee
DOCUMENT 10:35H830
0086-50 / ROD
DRAWING ID:
9M5471.CDR
DRAWING DATE:
FEBRUARY 10,1998 SB
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REACTOR
CONTROL
LEGEND
REACTOR VESSEL
RADIATOR STACK
MODIFIED FROM: ORNL DWG 63-1209R
2. HEAT EXCHANGER
3. FUEL PUMP
4. FREEZE FLANGE
5. THERMAL SHIELD
8. COOLANT PUMP
7. RADIATOR
8. COOLANT DRAIN TANK
0. FANS
10. DRAIN TANKS
11. FLUSH TANK
12. CONTAINMENT VESSEL
13. FREEZE VALVE
Fig. 2.3
Simplified MSRE flow diagram of primary
and secondary reactor circuits
DOE • ORNL. Molten Salt Reactor Experiment - Oak Ridge, Tennessee
DOCUMENT ID: 35M830
0086-50 / ROD
DRAWING ID:
97-15472.CDR
DRAWING DATE:
FEBRUARY 10. 1996 SB
preceeding blank page omitted
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tank head spaces and associated off-gas system. Fluorine generation was expected based on
knowledge about the chemical stability of fluoride salt. An annealing process was part of shut-
down procedures between 1971 and 1989. This process heated fuel salt to below melting
temperatures to force the fluorine in the salt matrix to recombine before it would migrate from
the salt. It appears that during the annealing process, unknown to operators, uranium
'hexafluoride gas was formed and liberated from the salt.
In 1994, investigation of the MSRE site indicated that anomalous levels of uranium
hexafluoride and fluorine gases were present throughout the off-gas piping connected to the fuel
and flush salt drain tanks. In addition, uranium had migrated through the off-gas system to an
auxiliary charcoal bed that resulted in a criticality concern because of the quantity of uranium
detected. Interim corrective measures were immediately taken to ensure the safety of workers
and personnel. Shortly afterwards, documentation of actions taken and continuing actions were
included in a CERCLA time-critical removal action memorandum. A plan was then developed
for remediating the MSRE site to reduce the risk presented by the continuing presence of the fuel
and flush salts in storage at MSRE. Planners organized mitigation of the migrated MSRE
uranium (as uranium hexafluoride) and fluorine gas into three separate CERCLA actions.
Time-Critical Removal Action. This CERCLA action, approved in July 1995
(DOE 1995), is completed. The interim corrective measures provided risk reduction for
employees and workers at MSRE by addressing various aspects of containment, nuclear criticality
control, and chemical reaction prevention. A reactive gas removal system, installed in 1996 as
part of the time-critical action, continues to remove and trap uranium hexafluoride and fluorine
gases from MSRE off-gas piping.
Non-Time-Critical Removal Action. Removal of the uranium deposit and associated
fluorine contaminated charcoal from the auxiliary charcoal bed was approved as a CERCLA non-
time-critical removal action (DOE 1996). Removal of uranium and fluorine contaminated
charcoal is planned for completion in February 1999. This action will eliminate the potential of
a criticality accident or chemical reaction in the charcoal bed cell and reduce the risk to human
health and environment from exposure to the toxic and radioactive uranium.
Remedial Action. This ROD for interim action focuses on removal of fuel and flush salts
from the MSRE drain tanks to eliminate the major source of contaminants for the MSRE site.
Potential sources of uranium hexafluoride and fluorine gases will be eliminated from the drain
tanks thereby reducing the risk to workers, employees, and the public. Contaminants that remain
at the MSRE site following this interim action and their associated risks will be addressed in a
JT00869709 1BH/CIE 2-8 June 3. 1998
-------
subsequent CERCLA action. The fuel and flush salts from MSRE will be treated to reduce risks
during storage while awaiting shipment for final disposition.
HIGHLIGHTS OF COMMUNITY PARTICIPATION
The interim action proposed plan for the MSRE site was released to the public in
December 1997. This document is part of the Administrative Record for this decontamination
and decommission action, which is maintained at the DOE Information Resource Center,
105 Broadway Avenue, Oak Ridge, Tennessee 37830. Notice of availability for this plan and
other documents in the Administrative Record was published in The Knoxville News-Sentinel
December 22, 1997, The Oak Ridger December 22, 1997, The Roane County News
December 24, 1997, and The Clinton Courier-News December 24, 1997. The public comment
period was held between December 23, 1997, and January 30, 1998. A public meeting held
January 14, 1998, to discuss the proposed plan resulted in verbal comments. Two written
comments were received during the public comment period. Responses to the written comments
and verbal comments from the public meeting relating to this interim action are presented in
Part 3, "Responsiveness Summary," of this document.
At the request of DOE, the National Research Council formed a committee of
distinguished scientists mid engineers in the spring of 1996 to review alternatives for removal and
disposition <£ MSBE fluoride salts. The first of two p&lifc meetings held by the committee
convened Sept&nber 9 and 10,1996, in Oak Ridge at the Garden Plaza Hotel, this meeting was
advertised in local newspapers and wa$ well attended fay the public. The second pubic meeting
was held October 3* 1996, i» Washington D,C~. to respond to tjuestjons previously raised by
panel members. In February 1997, the National Research Council released their report {NRC
1997). Recommendations made in the report are consistent with alternatives presented in the FS
and support the interim action approach recommended in the proposed plan and selected in this
ROD.
SCOPE AND ROLE OF THE SITE INTERIM REMEDIAL ACTION
The scope of this interim remedial action is to remove the fuel and flush salts from the
drain tanks, separate the uranium from the fuel and flush salts, convert the uranium to an oxide
for storage as part of the existing B3U repository inventory, stabilize/package the residual salt,
and place the residual salt in interim storage until an end-point location is selected for final
disposal. This interim action will eliminate the risk of a criticality incident and the hazards
JT00869709.IBH/CJE 2-9 J"1* 3- "98
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associated with uranium hexafluoride and fluorine gas release at the MSRE site. Decontamination
and demolition of Building 7503 and the MSRE reactor components will be performed as part
of a later, separate CERCLA final action. Ongoing management and final disposition of the
uranium oxide will be determined pursuant to the program for managing the existing 231U
repository inventory (rather than further CERCLA action).
SUMMARY OF SITE CHARACTERISTICS
This remedial action addresses the two contaminated waste salts at the MSRE site—fuel
salt and flush salt. The fuel and flush salts are stored in tanks in the drain tank cell below the
floor of Building 7503. The fuel salt is divided between two drain tanks, and the flush salt is
stored in one flush drain tank. All three tanks are similarly constructed; however, the fuel drain
tanks are equipped with steam domes and thimbles to remove heat produced by radioactive decay.
Heat production within the fuel salt is no longer a concern.
Both salts are composed of Li, Be, and Zr fluoride salts. The fuel and flush salts differ
in the amount of fuel and fission products contained in each, and the fuel salts have a higher
percentage of zirconium. The flush salt contains a small amount of the fuel and fission products
because it was used to flush residual fuel salt out of the reactor and the associated piping system
after the fuel salt was drained into the storage drain tanks. It is estimated that the flush salts
contain approximately 500 g (1.1 Ib) or 2.9 Ci of uranium and 13 g (< 0.1 Ib) or 1 Ci of
plutonium. Figure 2.4 describes the proportions of salts constituents at the end of reactor
operation. Table 2.1 lists the salt weight, volume, and density, and Table 2.2 lists the principal
isotopes in the salts after irradiation in the reactor. The mass of uranium in the fuel and flush
salts shown in Table 2.2 [approximately 37.5 kg (82 Ib)] represents the amount of uranium
[1.1 percent of the fluoride salts as uranium tetrafluoride (UF4)] that was transferred to the drain
tanks at the end of reactor operation. Since reactor shutdown, uranium has migrated from the
fuel salt to the drain tank head space, off-gas system, and an auxiliary charcoal bed in the form
of uranium hexafluoride. The current mass of uranium in the fuel salts is calculated to be
approximately 20 kg (44 Ib) (0.6 percent of the fluoride salts as UF4).
Fluorine liberation from the salts has left metallic Li, Be, and Zr in the salt and created
a net reducing condition in the salt. As a result the potential exists for uranium to precipitate
during the melting process. The present reducing potential of the stored salt is latent because the
metal is essentially immobile; however, once the salt is heated to melting temperatures, the
reduction reaction may proceed. During melting, the reducing potential could cause up to 12 kg
(26 Ib) of uranium metal to precipitate and/or diffuse into the tank wall. This could result in a
11-00869709.1BH/CJE 2-10 /unc 3. 1998
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Fuel Drain Tanks
2.479kg
No. 1
No. 2
2.171 kg
Fluoride Salts
E7LiF(42.3%)
• BeF2 (36.0%)
• ZrF4 (20.6%)
DUF4(1.1%)and
PuF3 (0.02%)
•OJU(0.02%)
and "HJ (0.1%)
Plutonium
•"•911(90.1%)
•J*Pu(9.50%)
•Other Pu (0.350%)
Note: Does not Include the -2.7kg
of fission products.
Fuel Flush Tank
4.265kg
Fluoride Salts
E37LiF(51.3%)
• BeF2 (47.8%)
• ZrF4 (0.89%)
DUF4 (0.015%)
PuF3 (0.0004%)
Uranium
•'"U (39.4%)
•"•0(3.6%)
•=^(17.4%)
o»U(39.4%)
•"U (0.008%)
Mo*, and »U (0.2%)
•z»Pu(94.7%)
•"•Pu(4.80%)
• Other Pu (0.500%)
Fig. 2.4
Composition of fuel and flush salts
by weight percent
DOE - ORNL. Molten Sat Reactor Experiment - Oak Ridge. Tennessee
DOCUMENT IDJ5HOO
0086-50/ROO
DRAWING ID:
S7-15«73jCOR
DRAWING DATE
FEBRUARY 17.1998 SR
2-11
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2. Compliance with Applicable or Relevant and Appropriate Requirements
On-site interim remedial actions under CERCLA are required to comply with only
those ARARs specific to the interim action being implemented.
Alternative 2 would not trigger any location-specific ARARs because this alternative
would not affect any sensitive resources. Water quality standards and Safe Drinking
Water Act maximum contaminant levels (MCLs) (which could be ARARs for the
groundwater and the springs during a final action) and other chemical-specific ARARs
are outside the scope of this interim action because no actions will be taken to alter
contamination levels. The final action for this site will be taken as part of the Upper
EFPC ROD, which will address Union Valley groundwater. MCLs will be ARARs
for setting cleanup goals for that action. Chapter 1200-l-13-.08(3)(a).(iv) of TDEC
final Rule, "Inactive Hazardous Substance Site Remedial Action Program." effective
February 19, 1994. requires institutional controls whenever a remedial action does not
address concentrations of hazardous substances that pose or may pose an unreasonable
threat to public health, safety, or die environment. This rule, however, is applicable
to actions "...consistent with a permanent remedy..." and is not applicable to this
interim action. Alternative 2 is an administrative remedy for an interim action and,
therefore, there are no location-, chemical-, or action-specific ARARs pertaining to
the proposed actions.
A statutory requirement under CERCLA [Sect. 121(b)(l)] requiring protection of
human health and the environment would not be met by the no action alternative
without some assurance that exposure pathways would remain incomplete in the future.
BALANCING CRITERIA
3. Long-Term Effectiveness and Permanence
For Alternative 2. long-term effectiveness is evaluated for the period beginning when
initial institutional controls (i.e., executing license agreements) are implemented per
this interim action ROD and ending when final remedial actions are implemented per
the Upper EFPC CA ROD. The interim actions include notification by property
owners of use or change of use of surface water or groundwater, prohibition of any
unacceptable actions, and annual title searches and notifications by DOE as a due-
diligence measure to identify undisclosed changes in ownership and remind owners of
their obligations. These actions are considered very effective for this interim period.
JT0004MI0.1MA/MBH 2-12 *«»* »• 1W7
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Table 2.1. Primary inventory of stored fuel and flush salts, MSRE site, ORNL,
Oak Ridge, Tennessee
Taflk
Salt weight
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Table 2.2. Activity of principal isotopes in the fuel and flush salts, MSRE site, ORNL, Oak Ridge, Tennessee
0
m
;:::S£ll;|
38
39
40
43
51
52
55
56
61
62
63
- ::::: j : ' : : : •'; •;• :•;•:•: • r-xj: •: •: '; • : • ' • .' '-''•''• £•:-;;":•:•:• :•; •: • :':; :-:-:>:-::
Strontium
Yttrium
Zirconium
Technetium
Antimony
Tellurium
Cesium
Barium
Promethium
Samarium
Europium
90
90
93
99
125
125
137
137/n
147
151
152
154
155
Total for fission
3i&&:£l£^
28.5 years
2.7 days
1.5 E6 years
2.1 E5 years
2.73 years
58 days
30 years
2.6m
2.62 years
90 years
13.3 years
8.8 years
4.96 years
products (2,711 g)
IlllllSllliiilll
7,550
7,550
0.3
0.5
1.0
0.3
6,290
5.940
50.3
121
1.5
4.7
9.3
27,500
CraMiunt isotaptf
92
Uranium
Total
232
233
234
for uranium
70 years
1. 59 E5 years
2.45 E5 years
isotopes (37,548 g)
135
302
17.4
454.4
%.:ffiBs::;.; ^:vm:"-:^ •;•.'. x- • : ••^.^ ^.Vf -.- ••:•::•••;. x^&^wp ;*;: .tossSivV. ••,;: :• ••<• • • : :,.:. • .-.•.••
:;:Aton»ic MixfHi^sym^r ::;;::i;::MaiM;;n<>. , -y-^:: -m$^:$yffi;i£; Activity ;(G)y : :•;•; :••;
81 Thallium
82 Lead
83 Bismuth
84 Polonium
85 Astatine
86 Radon
87 Francium
88 Radium
89 Actinium
90 Thorium
208
209
212
212
213
212
213
216
217
220
221
224
225
225
228
229
3.05 m
3.25 hours
10.6 hours
1 .01 hours
45.6m
45 seconds
4 ps
150ms
32ms
55.6 seconds
4.9m
3.66 days x
50
0.7
139
139
0.7
89.1
0.7
139
0.7
139
0.7
139
14.8 days A 0.7
10 days V« 0.7
1.9 days
7,300 years
Total for aclinide daughters (5.49 g)
94 Plutonium
95 Americium
Transuranium
238
239
240
241"
241
and olhtr iioivpti"
87.7 years
24, 110 years
6,540 years
14.4 years
433 years
Total for transuranics (737 g)
139
0.7
979
0.92
41.7
15.3
270
21.5
349.4
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Table 2.2. (continued)
Source: Table 6 of Williams, D. F., G. D. Del Cul, and L. M. Toth. 1996. A Descriptive Model of the Molten Salt Reactor Experiment After Shutdown.- Review ofFY 1995
Progress, ORNL/TM-13142. Oak Ridge National Laboratory, Chemical Technology Division, Oak Ridge, TN. The principal isotopes listed are those with a current
activity > 0.1 Ci. The total activity and weight for each isotope grouping includes other isotopes not listed here.
y
"Uranium and plutonium inventory values (except UJU) are derived from isotopic analysis and are 3 to 5 percent lower than those calculated by Bell, M. J. 1970. Calculated
Radioactivity of the Molten Salt Reactor Experiment Fuel Salt, ORNL/TM-2970. Oak Ridge National Laboratory, Oak Ridge, TN. All other projections are derived from the Bell
discharge inventory.
*Plutonium-241 is not a TRU waste element because its half-life is < 20 years.
Ci = curie ms = millisecond
g = gram MSRE = Molten Salt Reactor Experiment
> = greater than no. = number
< = less than ORNL = Oak Ridge National Laboratory
m = meter TRU = transuranic
MS = microsecond U = uranium
to
-------
SUMMARY OF SITE RISKS
Analysis shows that actual or threatened releases of hazardous substances from this site,
if not addressed by the preferred alternative or another active measure, present a current or
potential threat to public health, welfare, or the environment.
HUMAN HEALTH RISK
The streamlined risk assessment for the MSRE site evaluated two scenarios. A near-term
scenario postulates an exposure that could occur in the next 100 years after institutional controls
are lost. The other scenario postulates an exposure that could occur beyond 100 years. Included
on the risk assessment are only contaminants of potential concern with a credible exposure
pathway and long enough half-life to cause significant exposure if released. For the near-term
scenario, a release to the environment (air) from a failure in the off-gas piping connected to the
drain tanks was postulated. Contaminants of potential concern evaluated for this scenario
included fluorine gas, uranium hexafluoride gas, and HF gas. For the second scenario, a
criticality event was assumed to occur because of a failure in the drain tank cell and drain tanks.
Contaminants of potential concern were postulated as being fission-product gases generated by
a criticality event. Both scenarios evaluated the consequences to:
• an on-site receptor 100 m (328 ft) from the MSRE site and
• an off-site receptor 1,200 m (3,900 ft), the distance to the nearest public road, from
the MSRE site.
The exposure pathways quantified in this assessment were based on the conceptual site
model. The pathways included (1) a release of fluorine, uranium hexafluoride, and HF gases
because of an off-gas piping failure, which results in passerby exposure through the inhalation
and immersion pathways (near-term scenario) and (2) a criticality accident caused by a failure of
the drain tank cell and drain tanks resulting in passerby exposure from inhalation and immersion
in a cloud of radioactive gas (long-term scenario). No other exposure pathways were evaluated.
Based on EPA guidance for streamlined risk assessments, there is no need to evaluate all
»
pathways when risk is clearly exceeded by one exposure pathway.
The streamlined risk assessment showed that most of the estimated risks were above the
1 x 10" limit and were therefore unacceptable. For the near-term scenario, estimated risk for
JT00869709.IBH/CJE 2-16 iane 3- l998
-------
the on-site receptor is 5 x 10"' and ranges from 3 x 103 to 2 x 10~2 for the off-site receptor.
For the long-term scenario, the estimated risk for the criticality pathway is 1 x 102 for the on-
site receptor and 3 x 105 for the off-site receptor.
ECOLOGICAL RISK
The ecological risk assessment evaluated the potential for adverse effects on the
environment from exposure to contaminants in the MSRE drain tank cell. In the future, a
potential breach in a drain tank and a failure of the drain tank cell could contaminate groundwater
and surface water at nearby unnamed tributaries to White Oak Creek. The contaminated
groundwater would adversely affect terrestrial plants and wildlife. Thus failure of the fuel flush
tank or fuel drain tanks and the drain tank cell would adversely impact terrestrial plants and
wildlife. This scenario would also pose a risk to aquatic communities in nearby tributaries.
Aquatic receptors could be directly exposed by contact with and ingestion of contaminated water
and sediment. Terrestrial wildlife could also ingest contaminated surface water. Terrestrial flora
could be exposed to contaminated groundwater through root uptake.
DESCRIPTION OF ALTERNATIVES
An interim action alternative to reduce the risk posed by the fuel and flush salts at the
MSRE facility was developed and presented in the interim action proposed plan (DOE 1997a).
Use of this interim action will result in (1) reducing the risk at the MSRE facility and
(2) completing an action that is common in the alternatives that consider the ultimate disposition
of the salt for disposal.
The alternatives developed in the FS were prepared for an action that ideally would be
carried to completion with no delays. However, the locations identified in each alternative for
final salt disposition are currently not operational. Decisions about waste acceptance cannot be
made until locations for salt disposition are operational. As a result, none of the alternatives
developed in the FS can be fully implemented at this time. Selection of a disposal location for
MSRE salts must wait until one or both of the disposal facilities are opened and questions about
the acceptance of MSRE salts for disposal can be evaluated. In the interim, fuel and flush salts
will be removed from the MSRE facility. Uranium will also be removed from the salts and
managed as part of the existing 233U repository at ORNL. The salt remaining after the uranium
removal process will be stored until it is shipped to a disposal location.
JT00869709 1BH/CJE 2-17 June 3. 1998
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Five alternatives were developed in the FS to remove and dispose of the fuel and flush
salts (DOE 1997b). The alternatives consisted of a no further action alternative and four action
alternatives. The alternatives as presented in the FS are:
• Alternative 1: No Further Action,
• Alternative 2: Disposal at Waste Isolation Pilot Plant as Transuranic Waste,
• Alternative 3: Disposal at the National Repository as Spent Nuclear Fuel,
• Alternative 4: Disposal at the National Repository as High-Level Nuclear Waste, and
• Alternative 5: Disposal at a Combination of Sites as High-Level Nuclear Waste and
Low-Level Nuclear Waste.
The no further action alternative was evaluated as not meeting the purpose and the
objectives of this remedial action and therefore was not considered further. The four action
alternatives (Alternatives 2-5) each began by removing the salts from the MSRE facility and then
taking the actions necessary to transfer the salts to the designated end point for disposal. The
end-point locations for disposal of the salts or components of the salts are either the Waste
Isolation Pilot Plant (WIPP) in New Mexico as a defense-related transuranic (TRU) waste or a
national repository as either spent nuclear fuel or high-level nuclear waste. A decision now to
select a location for disposal of the MSRE salts could not be made with certainty that waste
acceptance criteria would be met. Evaluation and selection of a location for disposal of the
MSRE salt will be documented subsequently when an end-point location for disposal of the salt
is identified.
Another consideration for the MSRE site interim remedial action to remove salt from the
fuel and flush salt drain tanks is that removal can be completed without precluding the ultimate
disposal options. As indicated in each action alternative, removal of the fuel and flush salt from
the storage cell drain tanks is the first activity necessary for ultimate disposal of the salt. This
remedial action will include the salt in all three drain tanks, starting with the flush salt drain tank
which contains less radionuclides than either of the fuel salt drain tanks. Melting the salt in a
drain tank will start with a small volume and increase slowly .until all the salt is molten. To
chemically rebalance the salt, HF will be introduced into the molten salt as it melts. Uranium
will be separated from the molten salt using to the extent possible the same process and
equipment used to remove B5U in 1968. Fluorine gas will be added to the molten salt to oxidize
UF4 into uranium hexafluoride gas which will be trapped as it passes through vertical columns
JT00869709.IBH/CJE 2-18 June 3. 1998
-------
packed with sodium fluoride. The salt with the uranium removed will be moved from the drain
tanks into storage containers. The salt, which still contains a large quantity of radionuclides, will
then be stabilized/packaged to capture fluorine gas which may be generated. (The waste
containers will be placed in shielded casks for interim storage.) The casks will be set in an
existing storage facility at ORNL and managed there until final disposition is arranged.
INTERIM ACTION ALTERNATIVE
The MSRE interim remedial action activities are consistent with the FS salt disposal
alternatives. This action reduces risk and at the same time proceeds toward the end point of fuel
and flush salts disposal. Implementation of this interim action will not preclude any of the four
action alternatives from future consideration.
The ARARs developed in the FS have been reviewed and those pertinent to the interim
action are identified and presented in Tables 2.3 and 2.4.
SUMMARY OF COMPARATIVE ANALYSIS OF ALTERNATIVES
Implementation of the interim action would address the identified risks associated with
current conditions at the MSRE site. By separating uranium from the fuel and flush salts,
converting it to an oxide, packaging it in criticality-safe containers, and storing it in a facility
designed for the storage of a3U, risks associated with the release of uranium hexafluoride are
eliminated and risks of a nuclear criticality are managed in accordance with applicable standards.
By stabilizing/packaging the residual salt, fluorine gas generation can also be managed. This
action would allow DOE to defer decisions regarding further treatment and disposal of the salt
to a later date.
The comparative analysis using the nine CERCLA criteria for this interim remedial action
includes the no further action alternative and the interim action. Table 2.5 summarizes the
evaluation of the no further action alternative and this interim action (i.e., removal of salt,
separation of uranium, and interim storage of salt).
JT00869709.IBH/aE 2-19 June 3. 1998
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Table 2.3. ARARs for proposed activities, MSRE site, ORNL, Oak Ridge, Tennessee
ApplicaMlfty
0
X
Q
m
Alteration/destruction of
historic resources
Action(s) that will affect such resources must adhere to the DOE-ORO
Memorandum of Agreement (May 6, 1994). When alteration or
destruction of the resource is unavoidable, steps must be taken to
minimize or mitigate the impacts and to preserve data and records of
the resource
Any action that will impact historic
resources—applicable if there will
be alteration or modification
National Historic Preservation
Act of 1966 (16 USC 457a-w);
Executive Order 11593;
36 CFR 800;
DOE-ORO Programmatic
Agreement (May 5, 1994)
Chemical-speciflc
Release of radionuclides
during removal and storage
activities
N)
DOE will carry out all DOE activities to ensure that radiation dose to
individuals will be ALARA
Exposures to members of the public from all radiation sources shall not
cause an EDE to be > 100 mrem (1 mSv)/year
Management of TRU waste shall be conducted in such a manner as to
provide reasonable assurance that the combined annual dose equivalent
to any member of the public in the general environment resulting from
discharges of radionuclide material and direct radiation from such
management shall not exceed 25 mrem/year to the whole body and
75 mrem/year to any critical organ
Exposures to members of the public from all radiation sources released
into the atmosphere shall not cause an EDE to be > 10 mrem
(0.1 mSv)/year
Radiological emission measurements must be performed at all release
points that have a potential to discharge radionuclides into the air in
quantities which could cause an EDE in excess of 1 % of the standard
(0.1 mrem/year). All radionuclides which could contribute > 10% of
the standard (1 mrem/year) for the release point shall be measured
Release of radionuclides into the
environment—TBC
Handling and management of TRU
waste—relevant and
appropriate"*
Point source discharge of
radionuclides into the air from a
DOE facility—applicable
DOE Order 5400.5(1.4)
(proposed as 10 CFR 834)
DOE Order 5400.5(11,1 a)
(proposed as 10 CFR 834)
40 CFR 19I.03(b)
40 CFR 61.92;
Rules of the TDEC 1200-3-11-
.08 ^
40CFR61>93;
Rules of the TDEC 1200-3-11-
.08
-------
Table 2.3. (continued)
a
x
D
m
to
to
Characterization of TRU
TRU waste must be evaluated to determine the kinds and quantities of
TRU radionuclides present before storage
Radionudidc-contaminated
material; on-sile storage
Temporary storage of ftiel/
flush salts as a TRU waste
pending disposal
Interim storage/disposal of
LLW generated from the
separation process
(i.e., PPE, wipes.
contaminated hardware)
External exposures to the waste and concentrations of radioactive
material which may be released into the environment must not exceed
an EDE of 25 mrem/year to any member of the public
TRU waste shall be segregated or clearly identified to avoid
commingling of the waste with high-level, low-level waste or other
noncertified TRU waste
TRU waste storage areas must be protected from unauthorized access
TRU waste must be monitored periodically to ensure that wastes arc
not releasing their radioactive constituents
TRU waste storage areas must be designed, constructed, maintained,
and operated with a contingency plan to minimize the possibility of
fire, explosion, or accidental release of radioactive components
TRU waste storage areas must be operated in a way to maintain
radiation exposures to ALARA
Management of TRU waste shall be conducted in such a manner as to
provide reasonable assurance that the combined annual dose equivalent
resulting from discharges of radionuctide material and direct radiation
from such management shall not exceed 25 mrem/year to the whole
body and 75 mrem/year to any critical organ
Compliance with the pertinent WAC for the storage facility
Generation of TRU waste—TBC DOE Order 5820.2A (III.3b)
Storage of uranium after separation DOE Order 5820.2A (II.3a)
from salt—TBC
Temporary storage of TRU wastes DOE*Order 5820.2A (II.3.e)
at generating sites—TBC ' \
Handling and management of TRU 40 CFR 191,03(b)
waste—relevant and
appropriate**
Storage/disposal of LLW-TBC DOE Order 5820.2A (HI.3.e)
-------
Table 2.3. (continued)
5! "10 CFR 834.109 (proposed rule) requires that management of radioactive waste not exceed an EDE of 25 mrem/year from all exposure pathways. When prajnulgaied, this rule will
g be legally applicable. \
5 *DOE Order 5400.5, Chapter II l(c)(l), requires that TRU waste management and storage activities at facilities other than disposal facilities not cause members of the public to
g receive, in a year, a dose equivalent > 25 mrem to the whole body or a committed dose equivalent > 75 mrem to any organ.
m
ALARA = as low as reasonably achievable mSv = millisievert
ARAR = applicable or relevant and appropriate requirement ORNL. = Oak Ridge National Laboratory
CFR = Code of Federal Regulations ORO = Oak Ridge Operations
DOE = U.S. Department of Energy % = percent
EDE = effective dose equivalent PPE = personal protective equipment
> = greater than TBC = to be considered
< = less than TDEC = Tennessee Department of Environment and Conservation
LLW = low-level (radioactive) waste TRU = transuranic
mrem = millirem USC = United States Code
MSRE = Molten Salt Reactor Experiment WAC = waste acceptance criteria
to
SJ
to
-------
Table 2.4. Evaluation of the no further and preferred alternatives using the nine CERCLA criteria,
MSRE site, ORNL, Oak Ridge, Tennessee
Criteria
Overall protection of
human health and the
environment
Compliance with ARARs
Long-term effectiveness
Reduction of contaminant
toxicity. mobility, or
volume through treatment
Short-term effectiveness
Implementabiliry
Cost
State acceptance
Community acceptance
Evaluation
}"Jtt flatter BCtfamahernatfve ( Preferred alternative
THwhoM adttrfa
Poor. Existing controls will eventually be
inoperable and release of radioactive materials
from the salts would occur
Poor. Compliance over the long-term
questionable
Poor. Tanks containing salts will eventually fail
and release radioactive materials from the salts
Poor. Does not reduce toxicity, mobility or
volume through treatment
Good. The current controls collect uranium
hexafluoride and fluorine gases
Good. Reactive gas removal system in place
and operational
Poor. The present worth of operations and
maintenance for 70 years is $70 million to
maintain institutional and engineering controls
Good. Salts will be removed and placed in a safer, more stable configuration. This will reduce the
potential for an accidental release and allow for easier control of F: gases. The uranium fuel will be
separated and stored in an existing repository. This will eliminate generation of UF, gases
Yes. The proposed action complies with ARARs
Rolancbtg criteria
Good. Removes the principal threat from the MSRE facility by appropriately packaging the salts and
storing the packages in an appropriate facility. Removal of the salt is a permanent action
Good. Treatment to separate the uranium from the salts reduces toxicity of the salts and mobility is
reduced by converting uranium hexafluoride to uranium oxide. Volume is only incrementally reduced
because it is a small percentage of the total volume of the salt
Moderate. During activities of this alternative, risks from radiation and contamination exposure
associated with potential release will increase to workers and the public as the salt is heated, removed,
and containerized; however, safety analysis and appropriate precautions will be implemented to reduce
and control the risks
Moderate. The action is difficult yet feasible. Removal has been accomplished previously, but not
under current conditions. Interim storage will be at an existing storage facility at ORNL
Good. The total capital costs present worth of this action is $39.3 million
Madifytitg frittritt
The state of Tennessee and EPA are parties with DOE to the FFA and have considered this action as
presented in the feasibility study and proposed plan before approving this ROD
The interim action proposed plan was presented to the public for review between December 23, 1997,
and January 30, 1998, and no changes in the plans resulted based on the comments that were received.
Comments tended to support the proposed interim action. Stakeholders also participated in review of
the documents
to
ARAR = applicable or relevant and appropriate
requirement
CERCLA = Comprehensive Environmental
Response, Compensation, and Liability Act of 1980
$ = dollar
DOE = U.S. Department of Energy
EPA = U.S. Environmental Protection Agency
F2 = fluorine
FFA = Federal Facility Agreement
MSRE = Molten Salt Reactor Experiment
ORNL = Oak Ridge National Laboratory
ROD = record of decision
UF» = uranium hexafluoride
-------
Table 2.5. Estimated uranium in the salts before and after separation, MSRE site, ORNL,
Oak Ridge, Tennessee
Before uranium separation
Concentration
(jpjpw)
Mass
(kg)
Activity
{8C8/g)
After uranium separation
Concentration
(ppm)
Mass
Activity
(nCifg)
Fuel
233y
Total uranium
3,600
4,301
16.8
20
34,800
55,250
42
50
0.2
0.233
412
654
Flask
mv
Total uranium
46
117
0.2
0.5
450 .
673
20
50
0.08
0.214
192
289
g = gram
kg = kilogram
MSRE = Molten Sak Reactor Experiment
nCi = nanocurie
ORNL = Oak Ridge National Laboratory
ppm = parts per million
U = uranium
THE SELECTED REMEDY
The interim action remedy selected for the MSRE fuel and flush salts remediation is to
remove the salt in a chemically stable form, separate the uranium from the salts and store it
separately as part of the existing ^"U repository inventory, place the salt in containers, and store
the containerized salt until disposal is arranged. This action will employ the activities common
to the first steps in the removal and disposition of the fuel and flush salts for the four action
alternatives presented in the FS. The final action required for salt disposal will be documented
in a subsequent ||i|| CERCLA decision document and, as appropriate, in a NEPA decision
document.
Removal of salt from the drain tank cell will require new corrosive resistant equipment
to add heat and control the salt chemistry. To the extent possible, existing drain tanks and other
equipment will be examined and repaired for reuse, but requirements for operating the apparatus
remotely and adding HF to the melting salt exceed the original equipment capability. The goal
of the project is to remove 99 percent of the salts from each drain tank. This will reduce the
uranium mass left in each tank to below criticality safe limits.
JT00869709.IBH/CJE
2-24
June 3, 1998
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The separation of uranium from the fuel and flush salts will use the same process and, to
the extent practicable, the same equipment used to remove B5U in 1968. This process involves
adding fluorine to the molten salts. Uranium hexafluoride gas is liberated from the salts and then
trapped on vertical columns packed with sodium fluoride. The goal is to reduce the residual
uranium concentration in the salts to below 50 ppm. Depending on salt chemistry, it may be
possible to reproduce the results achieved in 1968 (26 ppm). Table 2.6 shows the estimated B3U
and total uranium concentrations before and after the separation process.
Uranium must be converted to uranium oxide to be placed in storage at the ORNL
repository. Although this conversion process is common in the uranium industry, a modification
tailored to a small scale, remote chemical operation will be applied to this application. The
chemically stable converted uranium will be packaged in suitable containers and prepared for
storage with similar packages in a B3U repository in Building 3019. Storage of this separated
uranium will result in approximately 17 kg (37 Ib) of B3U added to the 500 kg (1,100 Ib) of B3U
currently stored at the facility.
Once the uranium is separated from the salts, the residual salts will be poured into storage
containers (approximately 48 containers for the fuel and flush salt), and chemically stabilized/
packaged to capture fluorine gas which may be generated and to meet transportation requirements
for eventual shipment to a disposal area. Because a disposal facility is not available to make
waste acceptance determinations or to receive waste, the waste packages will be loaded into
shielded casks for interim storage. These casks will be placed in interim storage at an OMBl
operating storage facility. At present, facilities for remote handled waste include the RH-TRU
bunkers (Bldgs. 7&& and 7855), shielded storage well (e.g., 7827), and shielded concrete vaults
set on pads (e.g., 7842A)* inadequate and appropriate capacity does not exist in one of the
above facilities, a pad may becoostnicted or extended wi&in the existing boundaries of SW$A 5
or SWSA 6 specifically for the storage of MSKB salt residue waste caste. Final definition of the
shielded cask and storage site will be completed as part of the remedial design.
Total capital cost (present worth) to implement these interim activities is $39.3 million and
the annual operation and maintenance cost (present worth) are expected to be zero. The total
capital cost includes only the activities discussed in this section. Costs associated with interim
storage are not borne by this project; the $10,000 yearly costs are borne by other DOE-funded
programs. Other activities such as transportation to an end point disposal location identified in
the original four action alternatives are not included in this cost. Table 2.6 presents the schedule
»
for these activities.
Decisions concerning treatment and disposal of the salt is delayed to a later date. This
has the advantage that these decisions could be based on better information as waste acceptance
JT00869709.1BH/CJE 2-25 Iw* 3. 1998
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Table 2.6. Interim remedial action schedule, MSRE site, ORNL, Oak Ridge, Tennessee
Melt and transfer salts for processing
Separate uranium from salt
Transfer uranium to 233U repository
Stabilize and package salt
Interim storage of salts
Remedial action report
Start
July 2000
October 2000
October 2000
October 2000
October 2000
February 2003
Finish
May 2002
February 2003 s
February 2003 •-
February 2003
Undetermined
May 2003
Notes: Dales include operations. The durations do not include design, construction, etc.
MSRE = Molten Salt Reactor Experiment U = uranium
ORNL = Oak Ridge National Laboratory
criteria are developed and finalized for the national repository and WIPP, new treatment
technologies emerge, and further development is completed for existing treatment technologies
presented in the FS.
STATUTORY DETERMINATIONS
Section 121 of CERCLA establishes several statutory requirements and preferences,
including compliance with ARARs. CERCLA requires the remedy (1) be cost-effective; (2) be
protective of human health and the environment; (3) use permanent solutions and alternative
treatment technologies or resource recovery technologies to the maximum extent practicable; and
(4) use treatment that permanently reduces the toxicity, mobility, or volume of hazardous
substances. Interim remedial actions under CERCLA are required to attain only those ARARs
specific to the action being implemented, and the above criteria apply to the selection of a final
remedy. The selected interim action satisfies the above criteria.
This interim action provides short- and long-term protection of human health and the
environment through removal of a contaminant source and limitation of the potential spread of
contamination. This action will comply with al! ARARs. The action is cost-effective. The
action uses treatment to remove and stabilize uranium for storage in the B3U repository at ORNL
JT00869709.1BH/CJn
2-26
June 3, 1998
-------
and is permanent within the scope of the action because it removes the fuel and flush salts from
the MSRE facility. The proposed action also reduces the potential contaminant release and is
therefore appropriate as an interim action.
EXPLANATION OF SIGNIFICANT CHANGES
A review of all comments resulted in no significant changes to the remedy originally
identified in the proposed plan as the interim action alternative.
REFERENCES
DOE (U.S. Department of Energy). 1997a. Interim Action Proposed Plan for Fuel and Flush
Salt Disposition from the Molten Salt Reactor Experiment, Oak Ridge National
Laboratory, Oak Ridge, Tennessee, DOE/OR/02-1601&D3. Prepared by Jacobs
Environmental Management Team, Oak Ridge, TN.
DOE. 1997b. Feasibility Study for Fuel and Flush Salt Removal from the Molten Salt
Reactor Experiment at the Oak Ridge National Laboratory, Oak Ridge, Tennessee,
DOE/OR/02-1559&D2. Prepared by Jacobs EM Team, Oak Ridge, TN.
DOE. 1996. Action Memorandum for Uranium Deposit Removal at the Molten Salt Reactor
Experiment, Oak Ridge National Laboratory, Oak Ridge, Tennessee, DOE/OR/02-
1488&D2. Prepared by Jacobs EM Team, Oak Ridge, TN.
DOE. 1995. Oak Ridge National Laboratory Molten Salt Reactor Experiment Facility Time
Critical Removal Action Memorandum Report. Prepared by Lockheed Martin Energy
Systems, Inc., Oak Ridge, TN.
NRC (Natioaal Research Council).. 1997. Evaluation of the U.S, Deportment of Envy's
A&emafiveitfor the Remewd and Disposition, of Molten Salt Reactor Experiment
fluoride Sabs. National Academy f*ress, Washington, DC.
JT00869709.IBH/CJE 2-27 June 3. 1998
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RESPONSIVENESS SUMMARY
The Interim Action Proposed Plan for Fuel and Flush Salt Disposition from the Molten Salt
Reactor Experiment, Oak Ridge National Laboratory, Oak Ridge, Tennessee (DOE 1997a) was
released for public review December 22, 1997. The comment period for the public to consider
the alternatives developed for interim remediation of MSRE was announced in local newspapers
to begin December 23, 1997, and end January 30, 1998. The notice of availability for this plan
and other documents in the Administrative Record was published daily in The Knoxville News-
Sentinel and The Oak Ridger December 23, 1997, and biweekly and weekly in The Roane County
News and The Clinton Courier-News December 24, 1997. A public meeting was held in Oak
Ridge January 14, 1998. This public meeting was also announced in newspapers January 11 and
12, 1998.
Through newspaper announcements and other public relations efforts, DOE invited the
public to participate in the review of plans being recommended for interim remediation of MSRE.
The interim action proposed plan and other related documentation in the Administrative Record
were made available for review at the DOE Information Resource Center, 105 Broadway Avenue,
Oak Ridge, Tennessee. Written comments from the public could be received at the Information
Resource Center or sent to Ms. Margaret Wilson, DOE FFA Manager. DOE also accepted
written comments at the public meeting and responded to verbal comments. A transcript of the
public meeting is included in the Administrative Record.
DOE received two written comments during the public comment period. Responses to
these comments are included here. In addition, verbal comments that address the current
remedial action plan are included here to supplement the initial DOE response made at the public
meeting. Public comments and DOE responses that were made at the public meeting and which
do not address the plan for interim action are not included here.
LETTER 1
Comment: DOE and ORNL have approached the plan for MSRE fuel and flush salt
disposition in a thoughtful, forthright and honorable way.
»
Response: The support of the proposed plan is appreciated.
mM869709 IBH/CJE 3-3 J"« 3, 1998
preceeding blank page omitted
-------
LETTER 2
Comment: After review of the documents concerning the interim action proposed plan
for fuel and flush salt disposition and attending the public meeting, I fully concur with the
decision to select the preferred limited alternative which includes removal and interim
storage of the fuel and flush salts. I also studied the National Research Council report that
evaluated the alternatives for MSRE fuel and flush salts removal and disposition. This
report only solidified my opinion that the proposed plan was the correct one.
I was pleased that TDEC and EPA approved the proposed plan. I am concerned that
the regulatory process for approvals is not open to the public like the DOE decision process.
I would like to be part of the regulatory process to gain knowledge of their reasoning and
have the opportunity to discuss the reasons for decisions with the regulators.
Response: The support of the proposed plan is appreciated. Your desire for greater
involvement wifli TDEC and EPA has beea discussed with these agencies. The following,
provided by EPA, reaffirms support of public involvement and provides recommended avenues
to become involved in the CERCLA decision process.
The regulatory process for selecting CERCLA response actions ts open to the public.
TDEC and EPA «svi£» and comment on all documents prepared insoppor* of CBRCLAresponse
aclions. TDEC and EPA <#i*<^«de«ce is always avaSabJe to &e public. TDEC and EPA
participate in all formal public meetings and many information workshops. Additionally, TDEC
and EPA are represented on the Oak Ridge Ske-SpeciSc Advisory Board. Public involvement
in the regulatory process may be achieved through any of these means, as well as by direct oral
or written communications to TDEC and/or EPA representatives.
The public i$ an integral part of the regulatory process. Community acceptance of
response, action decisions is one of &e rate CERCtA remedy selection criteria that mast be
evaluated fojr all, remedial actions. However* 4$ regulatory agencies providing oversight af the
ORRy TDEC andEPA must consider DOE pop^sals indefendentty and then either concur or not
concur with those proposals, TDEC and EPA will provide me basis for their concurrence or
nonconcurrence and are available to discuss those decisions with the public.
JT00869709.IBH/CJE 3-4 June 3. 1998
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SUMMARY OF COMMENTS FROM THE PUBLIC MEETING
Comment 1: Three meeting participants commented that the proposed interim action
plan is appropriate and includes a reasonable approach for removing the salt from the
MSRE. In addition, even though the proposal does not include a recommendation for final
disposal of the salt, it is the correct action to take because it reduces thetisk of a release of
contaminants to the environment; and that the plan provided for due precautions to solve
a complex problem.
Response 1. The support of the proposed plan is appreciated.
Comment 2: Three meeting participants raised concerns about an alleged nuclear
criticality accident at the MSRE and alleged past releases/contamination incidents.
Response 2: Previous investigations determined that there has not been a criticality
accident at the MSRE, and that contamination incidents were minor and limited to two workers
in the facility. It is acknowledged, however, that there is the risk for a nuclear criticality accident
and substantial releases to the environment/public of fluorine gas and radioactive contamination
associated with the salts in the MSRE drain tanks. This is the reason that instead of the No
Action alternative, the proposed plan is to remove the salt from the drain tanks, remove the
uranium from the salt, stabilize/package the salt to control fluorine generation, and place the salt
containers in interim storage.
Comment 3: Suggestions for alternate remediation options were stated during the
public meeting by different commenters. These various options are presented with a brief
response.
(A) Has including the salt in the privatization initiative for transuranic waste
treatment after it is removed from MSRE been considered?
(B) Suggest melting the salt and placing it into containers for storage as spent nuclear
fuel. This would get it out of the way so you can go ahead and decontaminate and
decommission the MSRE building. But you will still have the fluorine problem wherever
you store the salt, and that may not be a job you want to do.
(C) Suggest fluorination to remove the uranium from the reactor and mix this
uranium with depleted uranium from K-25, denature the uranium, and make the uranium
JT00869709 IBH/CJE 3-5 >"tx 3. 1998
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safe. Then after that precipitate the uranium with either ammonia or sodium hydroxide and
make orange cake, and dispose of the orange cake in the burial grounds.
(D) [This idea was presented as not necessarily practical.] Suggest placing one or two
hundred tons of crushed limestone in the eel! (containing the fuel and flush salt storage
tanks) to fill it. That would take care of urdhium hexafluoride, excess fluorine, and
probably would take care of a rising water table.
Response 3:
(A) Yes, inquiries about including the MSRE salts in the privatization project have been
made; however, because the salts are unique in their chemical make-up with very little similarity
to other wastes at ORNL, inclusion of the salts is no longer considered.
(B) The suggestion to containerize and store the material as SNF implies not removing the
uranium before containerization. This was evaluated in the FS and discussed with the state of
Tennessee and EPA. It was determined that removing the uranium from the salt during the
current operations would be a small incremental cost to the project. Not removing the uranium,
however, may prevent future disposal at WIPP or prevent processing at INEEL for future
disposal at the National Repository. (Note: the work plan will address generation of fluorine
during interim storage.)
(C) The quantity of uranium (233U) that will be removed from the MSRE fuel and flush
salts is a very small amount compared with the quantity already stored in the B3U repository.
The process required to complete the suggested blending is not insignificant. Application of the
suggested process to address only the uranium from the fuel and flush salt would be inordinately
complicated and costly. The more appropriate implementation of this suggestion is to address
all of the a3U in the repository. Treatment of the repository inventory is beyond the scope of
this action.
(D) This interim remedial action is interim in part because it is only the first action for
the D&D of Building 7503, and this is the first action in removing, storage and disposition of the
fuel and flush salts. Before Building 7503 and MSRE can be decontaminated and
decommissioned, the fuel and flush salts must be removed. The salts cannot be left in place not
only because uranium hexafluoride and fluorine gases are liberated, but also because of the
hazards associated with and the regulatory guidance for disposition of spent nuclear fuel and/or
TRU waste. Leaving the fuel and flush salt in Building 7503 is not a viable option under these
circumstances, even if crushed limestone would be an effective temporary or permanent cover.
JT00869709.1BH/CJE 3-6 lunc 3- l998
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