The High Flux Isotope Reactor
| Since it began full-power operations
in 1966, the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory
(ORNL) has been one of the world's most powerful research reactors. The major
use of the HFIR is for neutron-scattering
experiments to reveal the structure and dynamics of a very wide range
of materials. The neutron-scattering instruments installed on the horizontal
beam tubes are used in fundamental studies of materials of interest to
solid-state physicists, chemists, biologists, polymer scientists, metallurgists,
and colloid scientists. These instruments are open to use by university
and industrial researchers on the basis of scientific merit.
One of the original primary purposes of the HFIR is the production of californium-252
and other transuranium isotopes for research, industrial, and medical applications.
These materials are produced in the flux trap in the center of the HFIR
fuel element where a working thermal-neutron flux of 2.0 x 1015
neutrons/(cm²·s) is available to irradiate the target material.
Additional irradiation facilities are also provided in the beryllium reflector.
Beyond its contributions to isotope production, the HFIR also provides for
a variety of irradiation tests and experiments that benefit from the exceptionally
high neutron flux available. In the fuel element flux trap, a hydraulic
rabbit tube provides access to the high thermal-neutron flux in the reactor
for short-term irradiations, and other positions are ideal for fast-neutron
irradiation-damage studies. A modification of the flux trap experiment facilities
in 1986 has provided two locations in the maximum flux region that can accommodate
instrumented capsules and engineering loops. The beryllium reflector contains
numerous experimental facilities with thermal-neutron fluxes up to 1.0 x
1015 neutrons/(cm²·s). These facilities can accommodate
static experimental capsules, complex fuel-testing engineering loops, and
special experimental isotope irradiations.
Each year about 150 to 200 researchers use the experiment facilities at
the HFIR. This booklet provides a general description of the HFIR, its capabilities,
and its accomplishments. It also provides potential users with descriptions
of the various individual experiment facilities.
The status of the transuranium production
program was critically reviewed by the U.S. Atomic Energy Commission (AEC)
Division of Research at a meeting on January 17, 1958. At that time the
AEC decided to embark on a program designed to meet the anticipated needs
for transuranium isotopes by undertaking certain irradiations in existing
reactors. By late 1958 it became apparent that acceleration of this program
was desirable. Following a meeting in Washington, D.C., on November 24,
1958, the AEC recommended that a high-flux reactor be designed, built, and
operated at ORNL, with construction to start in FY 1961.
As a result of this decision ORNL submitted a proposal to the AEC in March
1959. Authorization to proceed with the design of a high-flux reactor was
received in July 1959. The preliminary conceptual design of the reactor
was based on the "flux trap" principle, in which the reactor core consists
of an annular region of fuel surrounding an unfueled moderating region or
"island." Such a configuration permits fast neutrons leaking from the fuel
to be moderated in the island and thus produces a region of very high thermal-neutron
flux at the center of the island. This reservoir of thermalized neutrons
is "trapped" within the reactor, making it available for isotope production.
The large flux of neutrons in the reflector outside the fuel of such a reactor
may be tapped by extending empty "beam" tubes into the reflector, thus allowing
neutrons to be beamed into experiments outside the reactor shielding. Finally,
a variety of holes in the reflector may be provided in which to irradiate
materials for later retrieval.
In June 1961, preliminary construction activity was started at the site.
In early 1965, with construction complete, final hydraulic and mechanical
testing began. Criticality was achieved on August 25, 1965. The low-power
testing program was completed in January 1966, and operation cycles at 20,
50, 75, 90, and 100 MW began.
From the time it attained its design power of 100 MW in September 1966,
a little over 5 years from the beginning of its construction, until it was
temporarily shut down in late 1986, the HFIR achieved a record of operation
time unsurpassed by any other reactor in the United States. By December 1973,
it had completed its 100th fuel cycle, approximately 23 days each.
Notable accomplishments resulting from HFIR operation include the production
of californium-252, which is used for reactor startup sources, scanners
for measuring the fissile content of fuel rods, neutron activation analysis,
and fissile isotope safeguards measuring systems. In addition, californium-252
is used as a medical isotope to treat several types of cancer. Also, neutron
activation analysis at HFIR has been used by the semiconductor industry,
environmental remediation operations, and the Food and Drug Administration.
The Fusion Energy Program has been supported by the HFIR in three major
areas, including neutron interactive materials (structural materials and
ceramics), high heat flux materials, and plasma interactive materials.
The neutron-scattering facility
at HFIR has provided support to basic research programs involving neutron
scattering from polymers, colloids, magnetic materials, alloys, superconductors,
and biological materials.
In November 1986 tests on irradiation surveillance specimens indicated that
the reactor vessel was being embrittled by neutron irradiation at a rate
faster than predicted. The HFIR was shut down to allow for extensive reviews
and evaluation of the operation of this facility. Two years and five months
later, after thorough reevaluation, modifications to extend the life of
the plant while protecting the integrity of the pressure vessel, and upgrades
to management practices, the reactor was restarted. Coincident with physical
and procedural improvements were renewed training, safety analysis, and
quality assurance activities. Documents were updated, and new ones were
generated where necessary. Technical specifications were amended and reformatted
to keep abreast of the design changes as they were accepted by DOE. Not
only were the primary coolant pressure and core power reduced to preserve
vessel integrity while maintaining thermal margins, but long-term commitments
were made for technological and procedural upgrades.
After a thorough review of many aspects of HFIR operation, the reactor was
restarted for fuel cycle 288 on April 18, 1989, to operate initially at
very low power levels (8.5 MW) until all operating crews were fully trained
and it was possible to operate continuously at higher power. Following the
April 1989 restart, a further shutdown of nine months occurred as a consequence
of a question as to procedural adequacy. During this period, oversight of
the HFIR was transferred to the DOE Office of Nuclear Energy (NE); previously,
oversight was through the Office of Energy Research (ER). Following permission
by Secretary of Energy James Watkins to resume startup operation in January
1990, full power was reached on May 18, 1990. Ongoing programs have been
established for procedural and technological upgrade of the HFIR during
its operating life.
The HFIR is a versatile, 85-MW isotope production
and test reactor with the capability and facilities for performing a wide
variety of irradiation experiments. The HFIR is unique in the sense that
it provides one of the highest steady-state neutron fluxes available in
any of the world's reactors, and neutron currents from the four horizontal
beam tubes are among the highest available.
The original primary purpose of the HFIR was the production of transuranium
isotopes; however, many experiment-irradiation facilities were provided for
in the original design and several others have been added. Experiment-irradiation
facilities available include (1) four horizontal beam tubes, which originate
in the beryllium reflector; (2) the hydraulic tube facility, located in
the very high flux region of the flux trap, which allows for insertion and
removal of irradiation samples while the reactor is operating; (3) thirty
target positions in the flux trap, which normally contain transuranium production
rods but which can be used for the irradiation of other experiments (two
are instrumented target positions provided by a recent modification); (4)
six peripheral target positions located at the outer edge of the flux trap;
(5) numerous vertical irradiation facilities of various sizes located throughout
the beryllium reflector; (6) two pneumatic tube facilities in the beryllium
reflector, which allow for insertion and removal of irradiation samples while
the reactor is operating for activation analysis; and (7) four slant access
facilities, called "engineering facilities," located adjacent to the outer
edge of the beryllium reflector. In addition, spent fuel assemblies are
used for gamma irradiation in the gamma irradiation facility in the reactor
The HFIR is a beryllium-reflected, light-water-cooled and -moderated, flux-trap
type reactor that uses highly enriched uranium-235 as the fuel.
The reactor core assembly is contained in an 8-ft (2.44-m)-diam pressure
vessel located in a pool of water. The top of the pressure vessel is 17 ft
(5.18 m) below the pool surface, and the reactor horizontal midplane is 27.5
ft (8.38 m) below the pool surface. The control plate drive mechanisms are
located in a subpile room beneath the pressure vessel. These features provide
the necessary shielding for working above the reactor core and greatly facilitate
access to the pressure vessel, core, and reflector regions.
The reactor core consists of a series of concentric annular regions, each
approximately 2 ft (0.61 m) high. A 5-in. (12.70-cm)-diam hole, referred
to as the "flux trap," forms the center of the core. The target typically contains curium-244
and other transuranium isotopes and is positioned on the reactor vertical
axis within the flux trap. The fuel region is composed of two concentric
fuel elements. The inner element contains 171 fuel plates, and the outer
element contains 369 fuel plates. The fuel plates are curved in the shape
of an involute, thus providing a constant coolant channel width. The fuel
(U3O8-Al cermet) is nonuniformly distributed along
the arc of the involute to minimize the radial peak-to-average power density
ratio. A burnable poison (boron) is included in the inner fuel element primarily
to reduce the negative reactivity requirements of the control plates. The
average core lifetime with typical experiment loading is approximately 22
days at 85 MW.
The fuel region is surrounded by a concentric ring of beryllium reflector
approximately 1 ft (0.30 m) thick. This in turn is subdivided into three
regions: the removable reflector, the semipermanent reflector, and the permanent
beryllium is surrounded by a water reflector of effectively infinite thickness.
In the axial direction, the reactor is reflected by water.
The control plates, in the form of two thin, poison-bearing concentric cylinders,
are located in an annular region between the outer fuel element and the
beryllium reflector. These plates are driven in opposite directions. Reactivity
is increased by downward motion of the inner cylinder, which is used only
for shimming and regulation; that is, it has no fast safety function. The
outer control cylinder consists of four separate quadrants, each having an
independent drive and safety release mechanism. Reactivity is increased as
the outer plates are raised. All control plates have three axial regions
of different poison content designed to minimize the axial peak-to-average
power-density ratio throughout the core lifetime. Any single rod or cylinder
is capable of shutting the reactor down.
The reactor instrumentation and control system design reflects the emphasis
placed on the importance of continuity of operation while maintaining safe
operation. Three independent safety channels are arranged in a coincidence
system that requires agreement of two of the three for safety shutdowns.
This feature is complemented by an extensive "on-line" testing system that
permits the safety function of any one channel to be tested at any time during
operation. Additionally, three independent automatic control channels are
arrayed so that failure of a single channel will not significantly disturb
operation. All of these factors contribute to the continuity of operation
of the HFIR.
The primary coolant enters the pressure vessel through two 16-in. (40.64-cm)-diam
pipes above the core, passes through the core, and exits through an 18-in.
(45.72-cm)-diam pipe beneath the core. The flow rate is approximately 16,000
gpm (1.01 m³/s), of which approximately 13,000 gpm (0.82 m³/s)
flows through the fuel region. The remainder flows through the target, reflector,
and control regions. The system is designed to operate at a nominal inlet
pressure of 468 psig (3.33 x 106 Pa). Under these conditions
the inlet coolant temperature is 120°F (49°C), the corresponding
exit temperature is 156°F (69°C), and the pressure drop through
the core is about 110 psi (7.58 x 105 Pa).
From the reactor, the coolant flow is distributed to three of four identical
heat exchanger and circulation pump combinations, each located in a separate
cell adjacent to the reactor and storage pools. Each cell also contains
a letdown valve that controls the primary coolant pressure. A secondary
coolant system removes heat from the primary system and transfers it to the
atmosphere by passing water over a four-cell induced-draft cooling tower.
A graph showing an overview of the available
neutron fluxes in the HFIR is given in Fig. 1. Note that these are unperturbed
fluxes at 100 MW. Reduce the given values to 85% to account for the current
power level of 85 MW.
A fuel cycle for the HFIR normally consists of full-power operation at 85
MW for a period of from 21 to 23 days (depending on the experiment and radioisotope
load in the reactor), followed by an end-of-cycle outage for refueling. A
typical end-of-cycle refueling outage lasts approximately 4 to 6 days; however,
outages are occasionally extended as required to allow for control plate
changeout, calibrations, maintenance, and inspections. Experiment insertion
and removal may be accomplished during any end-of-cycle outage. Interruption
of a fuel cycle for experiment installation or removal is strongly discouraged.
Deviations from the schedule are infrequent and are usually caused by periodic
changeout of major reactor components, reactor and experiment component malfunctions,
The reactor has four horizontal beam tubes
which supply the neutrons to the instruments for the Center for Neutron Scattering.
Details for each beam tube and instrument can be found at the Center for Neutron Scattering website.
send comments or inquiries to Cindy Brackett.