[Federal Register Volume 87, Number 148 (Wednesday, August 3, 2022)]
[Notices]
[Pages 47400-47406]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2022-16573]


-----------------------------------------------------------------------

DEPARTMENT OF ENERGY

[DOE/EIS-0542]


Record of Decision for the Final Versatile Test Reactor 
Environmental Impact Statement

AGENCY: Idaho Operations Office, Department of Energy.

ACTION: Record of decision.

-----------------------------------------------------------------------

SUMMARY: The Department of Energy (DOE) is issuing this record of 
decision (ROD) for the Versatile Test Reactor (VTR) pursuant to the 
Final Versatile Test Reactor Environmental Impact Statement (VTR EIS) 
(DOE/EIS-0542). DOE prepared the VTR EIS to evaluate the potential 
environmental impacts of alternatives for constructing and operating a 
VTR and the associated facilities required for post-irradiation 
examination of test and experimental fuels and materials. DOE has 
decided to implement its Preferred Alternative, to construct and 
operate a VTR at the Idaho National Laboratory (INL) Site, and to 
establish, through modification and construction, co-located facilities 
for post-irradiation examination of test products and for management of 
spent VTR driver fuel at INL. The VTR will operate as a national user 
facility, providing a fast-neutron-spectrum test capability for the 
testing and development of advanced nuclear technologies. DOE has not 
decided whether to establish VTR driver fuel production capabilities at 
the INL Site, the Savannah River Site (SRS), or a combination of the 
two sites. Once a preferred alternative or option for VTR driver fuel 
production is identified, DOE will announce its preference in a Federal 
Register (FR) notice. DOE would then publish a ROD no sooner than 30 
days after its announcement of a preferred alternative/option for VTR 
driver fuel production.

ADDRESSES: Questions or comments should be sent to Mr. James Lovejoy, 
VTR EIS Document Manager, by mail at U.S. Department of Energy, Idaho 
Operations Office, 1955 Fremont Avenue, MS 1235, Idaho Falls, Idaho 
83415; or by email to [email protected]. The Final VTR EIS and 
this ROD are available for viewing or download at https://www.energy.gov/nepa/nepa-documents and https://www.energy.gov/ne/nuclear-reactor-technologies/versatile-test-reactor.

FOR FURTHER INFORMATION CONTACT: For information regarding the VTR 
Project, the Final VTR EIS, or the ROD, visit https://www.energy.gov/ne/nuclear-reactor-technologies/versatile-test-reactor; or contact Mr. 
James Lovejoy at the mailing address listed in ADDRESSES or via email 
at [email protected]; or call (208) 526-6805. For general 
information on DOE's National Environmental Policy Act (NEPA) process, 
contact Mr. Jason Anderson at the mailing address listed in ADDRESSES 
or via email at [email protected]; or call (208) 526-6805.

SUPPLEMENTARY INFORMATION:

Background

    DOE's mission includes advancing the energy, environmental, and 
nuclear security of the United States (U.S.) and promoting scientific 
and technological innovation in support of that mission. DOE's 2014 to 
2018 Strategic Plan states that DOE will ``support a more economically 
competitive, environmentally responsible, secure and resilient U.S. 
energy infrastructure.'' The plan further indicates that DOE will 
continue to explore advanced concepts in nuclear energy. The advanced 
concepts may lead to new types of reactors that improve safety, lower 
environmental impacts, and reduce proliferation concerns.
    Advanced reactors that operate in the fast-neutron \1\ spectrum 
offer the potential to have inherent safety characteristics 
incorporated into their designs. They can operate for long periods 
without refueling and reduce the volume of newly generated nuclear 
waste. Effective testing and development of advanced reactor 
technologies requires the use of fast neutrons comparable to those that 
would occur in actual advanced reactors. A high flux of fast neutrons 
allows accelerated testing, meaning that a comparatively short testing 
period would accomplish what would otherwise require many years to 
decades of exposure in a test environment with lower energy neutrons, a 
lower flux, or both. This accelerated testing would contribute to the 
development of materials and fuels for advanced reactors and generate 
data allowing advanced reactor developers, researchers, DOE, and 
regulatory agencies to improve performance, understand material 
properties, qualify improved materials and fuels, evaluate reliability, 
and ensure safety. Accelerated testing capabilities would also benefit 
these same areas for the current generation of light-water reactors.
---------------------------------------------------------------------------

    \1\ Fast neutrons are highly energetic neutrons (ranging from 
0.1 million to 10 million electron volts [MeV] and travelling at 
speeds of thousands to tens of thousands kilometers per second) 
emitted during fission. The fast-neutron spectrum refers to the 
range of energies associated with fast neutrons.
---------------------------------------------------------------------------

    Many commercial organizations and universities are pursuing 
advanced nuclear energy fuels, materials, and reactor designs that 
complement DOE and its laboratories' efforts to advance nuclear energy. 
These designs include thermal \2\ and fast-spectrum reactors that 
target improved fuel resource utilization and waste management, and the 
use of materials other than water for cooling. Their development 
requires an adequate infrastructure for experimentation, testing, 
design evolution, and component qualification. Available irradiation 
test capabilities are aging (most are over 50 years old). These 
capabilities are focused on testing materials, fuels, and components in 
the thermal neutron spectrum and do not have the ability to support the 
needs for fast reactors (i.e., reactors that operate

[[Page 47401]]

using fast neutrons). Only limited fast-neutron-spectrum testing 
capabilities, with restricted availability, exist outside the U.S.
---------------------------------------------------------------------------

    \2\ Thermal neutrons are neutrons that are less energetic than 
fast neutrons (generally, less than 0.25 electron volt and 
travelling at speeds of less than 5 kilometers per second), having 
been slowed by collisions with other materials such as water. The 
thermal neutron spectrum refers to the range of energies associated 
with thermal neutrons.
---------------------------------------------------------------------------

    A number of studies evaluating the needs and options for a fast-
neutron spectrum test reactor have been conducted. The Advanced 
Demonstration and Test Reactor Options Study identified a strategic 
objective to ``provide an irradiation test reactor to support 
development and qualification of fuels, materials, and other important 
components/items (e.g., control rods, instrumentation) of both thermal 
and fast neutron-based . . . advanced reactor systems.'' The DOE 
Nuclear Energy Advisory Committee (NEAC) issued an Assessment of 
Missions and Requirements for a New U.S. Test Reactor, confirming the 
need for fast-neutron testing capabilities in the U.S. and 
acknowledging that no such facility is readily available domestically 
or internationally. Developing the capability for large-scale testing, 
accelerated testing, and qualifying advanced nuclear fuels, materials, 
instrumentation, and sensors is essential for the U.S. to modernize its 
nuclear energy infrastructure and to develop transformational nuclear 
energy technologies that re-establish the U.S. as a world leader in 
nuclear technology commercialization.
    DOE's Mission Need Statement for the Versatile Test Reactor (VTR) 
Project, A Major Acquisition Project embraces the development of a 
well-instrumented, sodium-cooled, fast-neutron-spectrum test reactor in 
the 300 megawatt-thermal power level range. The deployment of a sodium-
cooled, fast-neutron-spectrum test reactor is consistent with the 
conclusions of the test reactor options study and the NEAC 
recommendation.
    As required by the Nuclear Energy Innovation Capabilities Act of 
2017 (NEICA) (Pub. L. 115-248), DOE assessed the mission need for a 
VTR-based fast-neutron source to serve as a national user facility. 
Having identified the need for the VTR, NEICA directs DOE ``to the 
maximum extent practicable, complete construction of, and approve the 
start of operations for, the user facility by not later than December 
31, 2025.'' The Energy Act of 2020, within the Consolidated 
Appropriations Act (Pub. L. 116-68), directs the Secretary of Energy to 
provide a fast-neutron testing capability, authorizes the necessary 
funding, and revises the completion date from 2025 to 2026. To this 
end, DOE prepared an EIS in accordance with NEPA and the Council on 
Environmental Quality (CEQ) and DOE NEPA regulations (40 CFR parts 1500 
through 1508 \3\ and 10 CFR part 1021, respectively).
---------------------------------------------------------------------------

    \3\ On July 16, 2020, the CEQ published an ``Update to the 
Regulations Implementing the Procedural Provisions of the National 
Environmental Policy Act'' (85 FR 43304). CEQ clarified that these 
regulations apply to NEPA processes begun after the effective date 
of September 14, 2020, and gave agencies the discretion to apply 
them to ongoing NEPA processes (40 CFR 1506.13). This VTR EIS was 
started prior to the effective date of the revised CEQ regulations, 
and DOE has elected to complete the EIS pursuant to the regulations 
in effect prior to September 14, 2020 (1978 regulations).
---------------------------------------------------------------------------

Purpose and Need for Agency Action

    The purpose of this DOE action is to establish a domestic, 
versatile, reactor-based fast-neutron source and associated facilities 
that meet identified user needs (e.g., providing a high neutron flux of 
at least 4 x 10 \15\ neutrons per square centimeter per second and 
related testing capabilities). Associated facilities include those for 
the preparation of VTR driver fuel and test/experimental fuels and 
materials and those for the ensuing examination of the test/
experimental fuels and materials; existing facilities would be used to 
the extent possible. The U.S. has not had a viable domestic fast-
neutron-spectrum testing capability for almost three decades. DOE needs 
to develop this capability to establish the U.S. testing capability for 
next-generation nuclear reactors--many of which require a fast-neutron 
spectrum for operation--thus enabling the U.S. to regain technology 
leadership for the next generation nuclear fuels, materials, and 
reactors. The lack of a versatile fast-neutron-spectrum testing 
capability is a significant national strategic risk affecting the 
ability of DOE to fulfill its mission to advance the energy, 
environmental, and nuclear security interests of the U.S. and promote 
scientific and technological innovation. This testing capability is 
essential for the U.S. to modernize its nuclear energy industry. 
Further, DOE needs to develop this capability on an accelerated 
schedule to avoid further delay in the U.S. ability to develop and 
deploy advanced nuclear energy technologies. If this capability is not 
available to U.S. innovators as soon as possible, the ongoing shift of 
nuclear technology dominance to other nations will accelerate, to the 
detriment of the U.S. nuclear industrial sector.

Proposed Action

    DOE proposes to construct and operate the VTR at a suitable DOE 
site. DOE would use or expand existing, co-located, post-irradiation 
examination capabilities as necessary to accomplish the mission. DOE 
would also use or expand existing facility capabilities to produce VTR 
driver fuel and to manage radioactive wastes and spent nuclear fuel. 
The DOE facilities would be capable of receiving test articles from the 
user community, as well as fabricating test articles for insertion in 
the VTR.\4\
---------------------------------------------------------------------------

    \4\ As a user facility, the VTR would provide experimental 
capabilities for entities outside of DOE. These other entities could 
also fabricate test items for placement in the reactor. The VTR 
project would develop procedures for the acceptance of test items 
for use in the VTR. All test item and assembly designs would be 
reviewed and verified to ensure that the VTR would perform as 
designed and would meet all core performance and safety requirements 
before the test assembly could be inserted into the reactor core.
---------------------------------------------------------------------------

    Candidate sites for construction and operation of the VTR include 
the INL Site near Idaho Falls, Idaho, and the Oak Ridge National 
Laboratory (ORNL), near Oak Ridge, Tennessee. DOE would perform most 
post-irradiation examination in existing, modified, or new facilities 
near the VTR, although there may be instances when test items would be 
sent to another location for evaluation. DOE would produce VTR driver 
fuel at the INL Site or SRS near Aiken, South Carolina.

Alternatives and Options Analyzed in the Final VTR EIS

    DOE proposes to use the GE Hitachi Nuclear Energy (GEH) Power 
Reactor Innovative Small Module (PRISM), a pool-type reactor, as the 
basis for VTR's design under both action alternatives. The PRISM design 
would require several changes, notably the elimination of electricity 
production and the accommodation for experimental locations within the 
core. The PRISM design \5\ of a sodium-cooled, pool-type reactor 
satisfies the need to use a mature technology. The VTR would be an 
approximately 300-megawatt (thermal) reactor based on and sharing many 
of the design and passive safety features of the GEH PRISM. It also 
would incorporate technologies adapted from previous sodium-cooled fast 
reactors (e.g., the Experimental Breeder Reactor II [EBR-II] and the 
Fast Flux Test Facility). The VTR's reactor, primary heat removal 
system, and safety systems would be similar to those of the PRISM 
design. VTR, like PRISM, would use

[[Page 47402]]

metallic alloy fuels. The conceptual design for the first VTR driver 
fuel core is an alloy of 70 percent uranium (uranium enriched to 5 
percent uranium-235 \6\), 20 percent plutonium, and 10 percent 
zirconium (by weight).
---------------------------------------------------------------------------

    \5\ The PRISM design is based on the EBR-II reactor, which 
operated for over 30 years. The PRISM design most like the VTR is 
the 471-megawatt thermal MOD-A design. The U.S. Nuclear Regulatory 
Commission review of the PRISM reactor, as documented in NUREG-1368, 
Preapplication Safety Evaluation Report for the Power Reactor 
Innovative Small Module (PRISM) Liquid-Metal Reactor, concluded that 
``no obvious impediments to licensing the PRISM design had been 
identified.''
    \6\ Enriched refers to the concentration of the isotope uranium-
235, usually expressed as a percentage, in a quantity of uranium. 
Low-enriched uranium (LEU), highly enriched uranium (HEU) and high 
assay, low-enriched uranium (HALEU) are all enriched forms of 
uranium. Depleted uranium is a byproduct of the enrichment process 
and refers to uranium in which the percentage of uranium-235 is less 
than occurs naturally.
---------------------------------------------------------------------------

    The major facilities in the VTR complex include an electrical 
switchyard, the reactor facility, 10 large sodium-to-air heat 
exchangers, and an operational support facility. The reactor facility 
would be about 180 feet by 280 feet. The reactor vessel, containing the 
core of the VTR, would extend 90 feet below grade. Other below-grade 
elements of the facility include the reactor head access area (over the 
core), secondary coolant equipment rooms, test assembly storage areas, 
and fuel cask pits. The reactor and experiment hall operating area that 
extends 90 feet above grade would allow the receipt and movement of 
fuel and experiments into and out of the core and storage areas.
    The VTR core design would differ from that of PRISM because it 
needs to meet the requirement for a high-flux test environment that 
accommodates several test and experimental assemblies. Experiments 
would be placed in some locations normally occupied by driver fuel in 
the PRISM. Heat generated by the VTR during operation would be 
dissipated through a heat rejection system consisting of intermediate 
heat exchangers within the reactor vessel, a secondary sodium-cooling 
loop, and air-cooled heat exchangers. This system and the Reactor 
Vessel Auxiliary Cooling System (RVACS) would provide shutdown and 
emergency cooling. The RVACS would remove decay heat from the sodium 
pool by transferring the thermal energy through the reactor vessel and 
guard vessel walls to naturally circulating air being drawn down 
through the inlets of four cooling chimneys, through risers on the 
exterior of the guard vessel, and up through the outlets of the cooling 
chimneys. The RVACS chimneys would be about 100 feet tall, extending 
above the experiment support area. No water would be used in either of 
the reactor cooling systems.
    The core of the VTR would comprise 66 driver fuel assemblies. The 
core would be surrounded by rows of reflector assemblies (114 total 
assemblies), which would be surrounded by rows of shield assemblies 
(114 total assemblies). Non-instrumented experiments (containing test 
specimens) could be placed in multiple locations in the reactor core or 
in the reflector region, by replacing a driver fuel or reflector 
assembly (test pins may also be placed within a driver fuel assembly). 
Instrumented experiments, which would provide real-time information 
while the reactor is operating, would require a penetration in the 
reactor cover for the instrumentation stalk and could only be placed in 
six fixed locations. One of these six locations can accommodate a 
``rabbit'' test apparatus that would allow samples to be inserted and/
or removed while the reactor is in operation. The number of 
instrumented test locations, plus the flexibility in the number and 
location of non-instrumented tests would strengthen the versatility of 
the reactor as a test facility.
    The VTR mission requires capabilities to examine the test specimens 
after irradiation in the VTR to determine the effects of a high flux of 
fast neutrons. Highly radioactive test specimens would be removed from 
the VTR after a period of irradiation ranging from days to years. Test 
specimens would then be transferred to a fully enclosed, radiation-
shielded facility where they could be remotely disassembled, analyzed, 
and evaluated. The examination facilities are ``hot cell'' facilities. 
These hot cells include concrete walls and multi-layered, leaded-glass 
windows several feet thick. Remote manipulators allow operators to 
perform a range of tasks on test specimens within the hot cell while 
protecting them from radiation exposure. An inert atmosphere is 
required in some hot cells. An inert atmosphere of argon would be used 
\7\ in the hot cell to which test assemblies are initially transferred 
after removal from the VTR. The inert atmosphere may be necessary to 
prevent test specimen degradation or unacceptable reactions (e.g., 
pyrophoric) that could occur in an air atmosphere. The post-irradiation 
hot cell facilities would be in close proximity to the VTR. After 
initial disassembly and examination in the inert atmosphere hot cell, 
test specimens may be transferred to other post-irradiation examination 
facilities for additional analysis.
---------------------------------------------------------------------------

    \7\ Not all test specimens would require an inert atmosphere 
during disassembly, analysis, and evaluation. However, separate 
facilities are not proposed for test specimens that do not require 
initial post-irradiation examination in an inert atmosphere.
---------------------------------------------------------------------------

    The VTR would generate up to 45 spent nuclear fuel assemblies per 
year.\8\ DOE would use existing or new facilities at the locations 
identified in the site-specific alternatives for the management of 
spent driver fuel. DOE will not separate, purify, or recover fissile 
material from VTR spent nuclear fuel. Spent driver fuel assemblies 
would be temporarily stored within the reactor vessel for about 1 year. 
Upon removal from the reactor vessel, surface sodium coolant would be 
washed off the assembly, and the assembly would be transported in a 
transfer cask to a new onsite spent fuel pad. After several years (at 
least 3 years), during which time the radioactive constituents would 
further decay, the assemblies would be transferred in a cask to a spent 
nuclear fuel conditioning facility. The sodium that was enclosed within 
the spent driver fuel pins to enhance heat transfer would be removed 
using a melt-distill-package process. The spent nuclear fuel would be 
chopped, and the chopped material consolidated, melted, and vacuum 
distilled to separate the sodium from the fuel. To meet safeguards 
requirements, diluent would be added to the remaining spent fuel to 
reduce the fissile material concentration. The resulting material would 
be packaged in containers and temporarily stored in casks on the spent 
fuel pad, pending transfer to an offsite storage or disposal facility. 
Currently, there is not a repository for disposal of spent nuclear 
fuel, but the conditioned spent driver fuel from the VTR is expected to 
be compatible with the acceptance criteria for any interim storage 
facility or permanent repository.
---------------------------------------------------------------------------

    \8\ Typically, less than a quarter of the VTR driver fuel 
assemblies would be replaced at the end of a test cycle. However, 
there could be atypical conditions when it would be necessary to 
replace a larger number of assemblies after a test cycle. In such 
instances, more than 45 assemblies could be removed from the core in 
a single year.
---------------------------------------------------------------------------

No Action Alternative

    Under the No Action Alternative, DOE would not pursue the 
construction and operation of a VTR. To the extent they are capable and 
available for testing in the fast-neutron-flux spectrum, DOE would 
continue to make use of the limited capabilities of existing 
facilities, both domestic and foreign. Domestic facilities that would 
likely be used, without modification, would include the INL Advanced 
Test Reactor and the ORNL High Flux Isotope Reactor. DOE would not 
construct new or modify any existing post-irradiation examination or 
spent nuclear fuel conditioning facilities to support VTR operation. 
Existing post-irradiation

[[Page 47403]]

examination and spent nuclear fuel conditioning facilities would 
continue to support operation of the existing reactors. Because there 
would not be a VTR under the No Action Alternative, there would be no 
need to produce VTR driver fuel. Therefore, no new VTR driver fuel 
production capabilities would be pursued. The No Action Alternative 
would not meet the purpose and need identified for the VTR.

Idaho National Laboratory Versatile Test Reactor Alternative

    Under the INL VTR Alternative, DOE would site the VTR adjacent to 
and east of the Materials and Fuels Complex (MFC) at the INL Site and 
use existing hot cell and other facilities at the MFC for post-
irradiation examination and conditioning spent nuclear fuel (i.e., 
preparing it for disposal). The VTR complex would occupy about 25 
acres. Additional land would be disturbed during the construction of 
the VTR complex for such items as temporary staging of VTR components, 
construction equipment, and worker parking. In total, construction 
activities (anticipated to last 51 months) would result in the 
disturbance of about 100 acres, inclusive of the 25 acres occupied by 
the completed VTR complex.
    The MFC is the location of the Hot Fuel Examination Facility 
(HFEF), the Irradiated Materials Characterization Laboratory (IMCL), 
and the Fuel Conditioning Facility (FCF). The HFEF and IMCL (and other 
analytical laboratory facilities) would be used for post-irradiation 
examination and the FCF for spent nuclear fuel conditioning. The 
existing Perimeter Intrusion Detection and Assessment System (PIDAS) 
security fencing around the Fuel Manufacturing Facility (FMF) and the 
Zero Power Physics Reactor (ZPPR) would be extended to encompass most 
of the VTR facility.
    Following irradiation, test and sample articles would be 
transferred to the HFEF first. The HFEF, a Hazard Category 2 nuclear 
facility,\9\ contains two large hot cells. HFEF hot cells provide 
shielding and containment for remote examination (including destructive 
and non-destructive testing), processing, and handling of highly 
radioactive materials.
---------------------------------------------------------------------------

    \9\ DOE defines hazard categories of nuclear facilities by the 
potential impacts identified by hazard analysis and has identified 
radiological limits (quantities of material present in a facility) 
corresponding to the hazard categories. Hazard Category 1--Hazard 
Analysis shows the potential for significant offsite consequences 
(reactors fall under this category). Hazard Category 2--Hazard 
Analysis shows the potential for significant onsite consequences 
beyond localized consequences. Hazard Category 3--Hazard Analysis 
shows the potential for only significant localized consequences. 
Below (Less Than) Hazard Category 3 applies to a nuclear facility 
containing radiological materials with a final hazard categorization 
less than Hazard Category 3 facility thresholds.
---------------------------------------------------------------------------

    The IMCL, a Hazard Category 2 nuclear facility, has a modular 
design that provides flexibility for future examination of nuclear fuel 
and materials. The IMCL would be used for the study and 
characterization of radioactive fuels and materials at the micro- and 
nanoscale to assess irradiation damage processes.
    Existing facilities within the MFC would need minor modifications 
to support fabrication of test articles or to support post-irradiation 
examination of irradiated test specimens withdrawn from the VTR. These 
types of activities are ongoing within the MFC.
    A new spent fuel pad would be constructed within the VTR site. The 
spent fuel pad would consist of an approximately 11,000-square foot 
concrete slab with a 2,500-square foot approach pad. Spent driver fuel 
would be temporarily stored at the VTR within the reactor vessel, 
followed by a period of storage on the spent fuel pad. After the fuel 
cools sufficiently, it would be transferred in a cask to FCF. FCF is a 
Hazard Category 2 nuclear facility located within a PIDAS. At FCF, the 
fuel would be conditioned using a melt-distill-package process. The 
fuel would be chopped, using existing equipment at the FCF. The chopped 
material would be consolidated, melted, and vacuum distilled to 
separate the sodium from the fuel. Following addition of a diluent, the 
mixture would be packaged in containers, placed in storage casks, and 
temporarily stored on the new spent fuel pad until shipped to an 
offsite location (an interim storage facility or a permanent repository 
when either becomes available for VTR fuel).
    Under the conceptual design, the existing infrastructure, including 
utilities and waste management facilities, would be used to support 
construction and operation of the VTR. The current infrastructure is 
adequate to support the VTR with minor upgrades and modifications. 
Radioactive wastes would be shipped off site for treatment and/or 
disposal.

Oak Ridge National Laboratory Versatile Test Reactor Alternative

    Under the ORNL VTR Alternative, the VTR would be sited at ORNL at a 
site previously considered for other projects, about a mile east of the 
ORNL main campus. The major structures for the VTR would be the same as 
those described for the INL VTR Alternative. At ORNL, a new hot cell, a 
joint post-irradiation examination and spent nuclear fuel conditioning 
facility, would be constructed adjacent to the VTR. Although there are 
facilities with hot cells at ORNL that would be used for post-
irradiation examination of test materials, none of the available hot 
cells operates with an inert atmosphere. A new spent fuel pad of the 
same dimensions as described under INL VTR Alternative would also be 
constructed.
    The new hot cell facility would be approximately 172 feet by 154 
feet, four levels, and would rise to about 84 feet above grade. The 
facility would house four hot cells: two for post-irradiation 
examinations and two for spent nuclear fuel conditioning. Construction 
would occur in parallel with the construction of the VTR and be 
completed in the same 51-month period. Construction activities would 
result in disturbance of about 150 acres, with the completed VTR 
complex, including the hot cell facility, occupying less than 50 acres. 
The VTR facility, hot cell facility, and spent fuel pad would be 
located within a single PIDAS.
    In addition to the new hot cell facility, existing facilities at 
ORNL within the Irradiated Fuels Examination Laboratory (Building 3525) 
and the Irradiated Material Examination and Testing Facility (Building 
3025E) would be used to supplement the capabilities of the new post-
irradiation examination facility. The Irradiated Fuels Examination 
Laboratory is a Hazard Category 2 nuclear facility and contains hot 
cells that are used for examination of a wide variety of fuels. The 
Irradiated Material Examination and Testing Facility is a Hazard 
Category 3 nuclear facility and contains hot cells that are used for 
mechanical testing and examination of highly irradiated structural 
alloys and ceramics. In addition, the Low Activation Materials Design 
and Analysis Laboratory would be used for the examination of materials 
with low radiological content that do not require remote manipulation.
    Spent driver fuel would be managed the same as described under the 
INL VTR Alternative--temporarily stored at the VTR reactor vessel, 
stored on the spent fuel pad, then conditioned and packaged. 
Conditioning spent nuclear fuel in preparation for disposal would occur 
in an inert atmosphere hot cell located in the new hot cell facility 
adjacent to VTR. Containerized spent nuclear fuel would be placed in 
storage casks and temporarily stored on the new spent fuel pad until 
shipped to an offsite location (an interim storage facility or a 
permanent repository when either becomes available for VTR fuel).

[[Page 47404]]

    Under the conceptual design, the existing ORNL infrastructure would 
be extended to the VTR site. The location selected for the VTR is 
relatively undeveloped and does not have sufficient infrastructure 
(e.g., roads, utilities, security) to support construction and 
operation of the VTR. Radioactive waste would be shipped off site for 
treatment and/or disposal. Waste management capabilities provided by 
the project (e.g., treatment or packaging of radioactive liquid waste) 
and facilities within ORNL would be used to support waste management 
during construction and operation of the VTR.

Reactor Fuel Production Options

    The VTR design envisions the use of metallic fuel. The initial VTR 
core would consist of a uranium/plutonium/zirconium alloy (U/Pu/Zr) 
fuel that would be 70 percent uranium (uranium enriched to 5 percent 
uranium-235), 20 percent plutonium, and 10 percent zirconium--a blend 
identified as U-20Pu-10Zr. VTR driver fuel used in later operations 
could consist of these elements in different ratios and could use 
plutonium with uranium of varying enrichments, including depleted 
uranium or uranium enriched up to 19.75 percent. Annual heavy metal 
requirements would be approximately 1.8 metric tons of fuel material 
(between 1.3 metric tons and 1.4 metric tons of uranium and between 0.4 
and 0.54 metric tons of plutonium, depending on the ratio of uranium to 
plutonium).\10\ Feedstock for this fuel could be acquired from several 
existing sources.
---------------------------------------------------------------------------

    \10\ The cited quantities are those for finished fuel as it is 
placed in the reactor and correspond to fuel that is from 20 to 27 
percent plutonium. Accounting for additional material that ends up 
in the waste during the reactor fuel production process, up to 34 
metric tons of plutonium could be needed for startup and 60 years of 
VTR operation.
---------------------------------------------------------------------------

    DOE's plan for providing uranium for fabricating VTR driver fuel is 
to acquire metallic uranium from a domestic commercial supplier. If 
another source of uranium were to be selected, DOE would conduct a 
review to determine if additional NEPA analysis would be needed. Other 
possible sources are DOE managed inventories of excess uranium acquired 
from many sources, including U.S. defense programs and the former DOE 
uranium enrichment enterprise. Some of the uranium is enriched and 
could be down-blended for use in VTR driver fuel.
    Existing sources of U.S. excess plutonium \11\ managed by DOE and 
the National Nuclear Security Administration (NNSA) would be sufficient 
to meet the needs of the VTR project. Potential DOE/NNSA plutonium 
materials include surplus pit \12\ plutonium (i.e., metal), other 
plutonium metal, oxide, and plutonium from other sources. If the U.S. 
sources cannot be made available for the VTR project or to supplement 
the domestic supply, DOE has identified potential sources of plutonium 
in Europe.
---------------------------------------------------------------------------

    \11\ Excess plutonium includes pit and non-pit plutonium that is 
no longer needed for U.S. national security purposes.
    \12\ A pit is the central core of a primary assembly in a 
nuclear weapon and is typically composed of plutonium metal (mostly 
plutonium-239), enriched uranium, or both, and other materials.
---------------------------------------------------------------------------

    VTR driver fuel production evaluated in the EIS involves two steps 
or phases: feedstock preparation and fuel fabrication. Depending on the 
impurities of the source material, a polishing process, or a 
combination of processes, would be required. These processes would be 
performed in a series of gloveboxes \13\ to limit worker radiological 
exposure.
---------------------------------------------------------------------------

    \13\ Gloveboxes are sealed enclosures with gloves that allow an 
operator to manipulate materials and perform other tasks while 
keeping the enclosed material contained. In some cases, remote 
manipulators may be installed in place of gloves. The gloves, glass, 
and siding material of the glovebox are designed to protect workers 
from radiation contamination and exposure.
---------------------------------------------------------------------------

    Three potential feedstock preparation processes are under 
consideration: an aqueous capability, a pyrochemical capability, and a 
combination of the two. In the aqueous process, the plutonium feed 
(containing impurities) is dissolved in a nitric acid solution and 
through a series of extraction and precipitation steps, a polished 
plutonium oxide is produced. The oxide is converted to a metal in a 
direct oxide reduction process. In one form of the pyrochemical process 
(molten salt extraction), the metallic plutonium feed is combined with 
a salt and the mixture raised to the melting point. Impurities (e.g., 
americium) react with the salt, and the polished plutonium is collected 
at the bottom of the reaction crucible. If the pyrochemical process 
were selected, a direct oxide reduction process would also be required 
to convert plutonium dioxide feeds to plutonium metal. If a combination 
of the two processes were to be selected, a smaller aqueous line to 
prepare this fuel could be incorporated into the pyrochemical process.
    Fuel fabrication would use an injection casting process to combine 
and convert the metallic ingots into fuel slugs. In a glovebox, a 
casting furnace would be used to melt and blend the three fuel 
components: uranium, plutonium, and zirconium. The molten alloy then 
would be injected into quartz fuel slug molds. After cooling, the molds 
would be broken, and the fuel slugs retrieved. Fuel pins would be 
created, using stainless steel tubes (cladding) into which a slug of 
solid sodium would be inserted, followed by the alloy fuel slugs. The 
fuel slugs and sodium would occupy about half of the volume of the fuel 
pin with the remainder containing argon gas at near atmospheric 
pressure. The ends of the tubes would be closed with top and bottom end 
plugs. These activities would take place in gloveboxes with inert 
atmospheres. Once fully assembled, the fuel pins would be heated 
sufficiently to melt the sodium and create the sodium bond with the 
fuel. The sodium-bonded fuel would fill about half the length of the 
fuel pin. Fuel pins would be assembled into a fuel assembly with each 
fuel assembly containing 217 fuel pins. Sodium bonding and producing 
the fuel assemblies would be performed in an open environment. No 
gloveboxes would be required.
    Operationally, the feedstock preparation and fuel fabrication 
capabilities would need to generate about 66 fuel assemblies for the 
initial VTR core. Thereafter, the capabilities would need to produce up 
to 45 fuel assemblies per year.
    The EIS evaluates the INL Site and SRS as potential locations for 
performing the activities necessary for driver fuel production for the 
VTR. Independently, DOE would establish and operate all or part of the 
fuel fabrication capability at either site. DOE is not making a 
decision regarding driver fuel production in this ROD.

Potential Environmental Impacts

    Implementation of either the INL VTR Alternative or the ORNL VTR 
Alternative would generally have small environmental consequences. 
Overall, the environmental consequences would be smaller at the INL 
Site for several reasons. The total area that would be temporarily 
disturbed and the area that would be permanently occupied by the VTR 
complex would be smaller at the INL Site because of the need to build a 
new hot cell facility if the VTR were located at ORNL. Unlike the INL 
Site, the ORNL location abuts wetlands that would have to be avoided or 
managed in accordance with Clean Water Act and State of Tennessee 
regulations. The removal of trees at the ORNL location would also 
result in the loss of roosting habitat for sensitive bat species. The 
potential radiological impacts would be small at both locations but 
would be smaller at the INL Site because the VTR would be further from 
the site boundary and the population density is lower near the INL Site 
than near ORNL.

[[Page 47405]]

    Implementation of the reactor fuel production options at either the 
INL Site or SRS would generally have small environmental consequences. 
At both locations, existing facilities would be modified or adapted to 
provide capabilities for feedstock preparation and fuel fabrication. 
Disturbance of a minimal area (up to 3 acres) would occur at SRS. 
Because there is existing staff at the INL Fuel Manufacturing Facility, 
fewer new employees would need to be hired for fuel fabrication at the 
INL Site. Potential radiological impacts would be small at both sites, 
but due to differences in population density and distribution, 
potential impacts would be somewhat smaller at the INL Site.

Environmentally Preferable Alternative

    The No Action Alternative would be the Environmentally Preferable 
Alternative. Under the No Action Alternative, DOE would not pursue the 
construction and operation of a VTR. To the extent they are capable and 
available for testing in the fast-neutron-flux spectrum, DOE would 
continue to make use of the limited capabilities of existing 
facilities, both domestic and foreign. Construction and operation of a 
VTR and associated support facilities would not occur, resulting in 
less impacts than under the Action Alternatives. However, the No Action 
Alternative would not meet the purpose and need for a domestic fast-
neutron-spectrum testing capability.

Comments on Final VTR EIS

    DOE made more than 1,850 notifications of the completion and 
availability of the Final VTR EIS to Congressional members and 
committees; states, including Idaho, Tennessee, and South Carolina; 
Tribal governments and organizations; local governments; other Federal 
agencies; non-governmental organizations; and individuals. Following 
issuance of the Final VTR EIS, DOE received four letters and/or emails. 
DOE considered the comments received following issuance of the Final 
VTR EIS and finds that they do not present ``significant new 
circumstances or information relevant to environmental concerns and 
bearing on the proposed action or its impacts'' within the meaning of 
40 CFR 1502.9(c) and 10 CFR 1021.314(a), and therefore do not require 
preparation of a supplement analysis or a supplemental EIS.
    DOE addressed two of the emails received--a press inquiry and a 
process question--directly with the people who submitted them.
    A third email/letter received included multiple comments on a 
variety of topics. One related to the author's Freedom of Information 
Act request and has no bearing on or relevance to the environmental 
impacts evaluated in the EIS. It also contained another question of 
whether the Office of Nuclear Energy would have the ability and funds 
to establish a VTR fuel fabrication project at SRS. As appropriate, the 
VTR EIS evaluated the potential environmental impacts of a fuel 
fabrication capability at SRS; the administrative and funding items are 
factors DOE would consider when it makes a decision regarding fuel 
fabrication.
    Other comments posed questions about the plutonium for VTR driver 
fuel fabrication, a nonproliferation assessment, and management of 
transuranic waste resulting from fuel fabrication activities. Similar 
topics were raised in comments on the Draft VTR EIS. DOE responded to 
these comment topics in Volume 3 of the Final VTR EIS and revised the 
EIS as necessary to fully address these topics commensurate with the 
stage of project development.
    This third letter/email also incorrectly stated that the VTR had 
been ``terminated'' and the ``EIS [was] improperly issued after 
termination.'' Additionally, it requested ``that no Record of Decision 
(ROD) be issued on the project.'' While it is correct that Congress did 
not appropriate funds for VTR in fiscal year 2022, the Energy Act of 
2020, included in the Consolidated Appropriations Act (Pub. L. 116-68), 
authorized full funding for the VTR project. DOE is following Council 
on Environmental Quality guidance to integrate NEPA into the planning 
process early to ensure planning and decisions reflect environmental 
values, to avoid delays, and to head off potential conflicts. By 
issuing the Final VTR EIS and ROD, DOE is taking important steps, 
consistent with the Energy Act of 2020, by deciding whether and where 
to construct the VTR. In accordance with its authorization in the 
Energy Act of 2020, DOE will work with Congress to obtain the funding 
needed to execute this important project.
    The fourth letter/email recommended that DOE clarify management 
approaches for spent driver fuel beyond January 1, 2035. As indicated 
in the response to comments received from the State of Idaho and as 
revised in the Final VTR EIS, prior to issuing this ROD, DOE committed 
to exploring potential approaches with the State of Idaho to clarify 
and, as appropriate, address potential issues concerning management of 
VTR spent nuclear fuel beyond January 1, 2035; those discussions are 
ongoing. Spent driver fuel from the VTR, regardless of whether it was 
generated before or after January 1, 2035, would be stored within the 
VTR reactor vessel until decay heat generation is reduced to a level 
that would allow fuel transfer and storage of the fuel assemblies with 
passive cooling. After allowing time for additional radioactive decay, 
the spent fuel would be transferred to a spent nuclear fuel 
conditioning facility. At the facility, the spent fuel would be 
chopped, melted, and vacuum distilled to remove the sodium, after which 
the fuel would be diluted and placed in canisters ready for future 
disposal. The canisters would be placed in dry storage casks and stored 
on site in compliance with all regulatory requirements and agreements. 
This VTR spent nuclear fuel would be managed at the site until it is 
transported off site to an interim storage facility or a permanent 
repository.

Decision

    DOE has decided to implement its Preferred Alternative as described 
in the Final VTR EIS. DOE's Preferred Alternative is to construct and 
operate a VTR at INL, and to establish, through modification and 
construction, co-located facilities for post-irradiation examination of 
test products and for management of spent VTR driver fuel at INL.
    DOE has not decided whether to establish VTR driver fuel production 
capabilities for feedstock preparation and fuel fabrication at the INL 
Site, SRS, or a combination of the two sites. Once a preferred 
alternative/option for VTR driver fuel production is identified, DOE 
will announce its preference in an FR notice. DOE would publish a 
record of decision no sooner than 30 days after its announcement of a 
preferred alternative/option for VTR driver fuel production.

Basis for the Decision

    The Final VTR EIS provided the DOE decision-maker with important 
information regarding potential environmental impacts of alternatives 
and options for satisfying the purpose and need. In addition to 
environmental information, DOE considered other factors including 
public comments, statutory responsibilities, strategic objectives, 
technology needs, safeguards and security, cost, and schedule, when 
making its decision.

Mitigation Measures

    No potential adverse impacts were identified that would require 
additional

[[Page 47406]]

mitigation measures beyond those required by regulation and agreements 
or achieved through design features or best management practices. 
However, the INL VTR Alternative has the potential to affect one or 
more resource areas. If during implementation, mitigation measures 
above and beyond those required by regulations are identified to reduce 
impacts, they would be developed, documented, and executed.

Signing Authority

    This document of the Department of Energy was signed on July 22, 
2022, by Robert Boston, Manager, Idaho Operations Office, Office of 
Nuclear Energy, pursuant to delegated authority from the Secretary of 
Energy. That document with the original signature and date is 
maintained by DOE. For administrative purposes only, and in compliance 
with the requirements of the Office of the Federal Register, the 
undersigned DOE Federal Register Liaison Officer has been authorized to 
sign and submit the document in electronic format for publication, as 
an official document of the Department of Energy. The administrative 
process in no way alters the legal effect of this document upon 
publication in the Federal Register.

    Signed in Washington, DC, on July 29, 2022.
Treena V. Garrett,
Federal Register Liaison Officer, U.S. Department of Energy.
[FR Doc. 2022-16573 Filed 8-2-22; 8:45 am]
BILLING CODE 6450-01-P