99mtc production in north america
TRANSCRIPT
-
7/31/2019 99mTc Production in North America
1/34
99m
Tc PRODUCTION PROCESSES: AN EXAMINATION OF PROPOSALS TOENSURE STABLE NORTH AMERICAN MEDICAL SUPPLIES
Submitted by
Margaret Cervera
Department of Environmental and Radiological Health Science
In partial fulfillment of the requirements
for the Degree of Master of Science
Colorado State University
Fort Collins, Colorado
Spring 2009
-
7/31/2019 99mTc Production in North America
2/34
Introduction
As a nation we are fearful of things we do not understand and producing a radioactive
medical treatment from material used to create nuclear weapons is little understood and
therefore generates great fear. The United States (US) currently relies on technetium-99
metastable (99mTc) for a large number of medical diagnostic procedures with an ever
increasing need, especially as the population ages. Supplies of99m
Tc reach the United
States via Canadas National Research Universal (NRU) reactor which is the sole large
scale molybdenum-99 (99
Mo) production facility in North America. This process
currently depends upon multiple stages of cooperation that are susceptible to disruption at
any time. The production of99mTc begins with shipping uranium-235 (235U) to Canada,
irradiation of manufactured targets in a high neutron flux, removal of99
Mo fission
products from the reactor and radiochemical separation of the 99Mo. The 99Mo-99mTc
generator can then be manufactured, shipped to a radiopharmacy and finally99m
Tc is
eluted for mixture to a final dosage form. Many agencies realize the urgent need for
domestic production of99mTc but the ability to implement currently approved processes
do not exist and changing the manufacturing process is a complex scientific and
regulatory issue.
For many people radioactivity in any form conjures visions of movie monsters,
nuclear accidents such as Chernobyl and Three Mile Island and nuclear war. Society
hears that weapons grade high enriched uranium (HEU) is being shipped around and out
of the country and they are concerned; What if it is stolen? Will domestic fissile materials
be used against us in an attack reminiscent of the Oklahoma City bombing or
September 11th attacks? Will citizens be exposed to deadly radiation and require
2
-
7/31/2019 99mTc Production in North America
3/34
evacuation from their homes? Will families face cancer as a result of any accidents
involving transport or nuclear material? Where is the waste going to be stored and will it
leak into my drinking water? In many ways these are valid concerns since accidents and
acts of negligence have occurred with radioactive materials although generally with more
limited consequences than many people believe. The technical aspects of generating
99Mo, or any other radioisotope used for medical or research applications will be a
challenge to present to the public in a way that will answer to concerns but not confuse
people further. To a certain extent the publics concerns can be alleviated by education
and openness. The purpose of this paper is to provide an insight into, and comparison of,
the current state of production, the possible changes that are being considered and the
infrastructure and scientific issues that will need to be resolved for the United States to
supply this vital medical isotope.
Importance and Use of99m
Tc
Tc-99m is the most commonly used medical isotope for medical imaging. It is
involved in about 70% of all nuclear medicine procedures. Activities, in dosage form, for
one brand, TechneLite, can range from just over 2 MBq/kg (0.05 mCi/kg) for pediatric
thyroid imaging to 1110 MBq (30 mCi) for blood pool imaging of the heart (Bristol-
Myers Squibb 2005). One of the highest use applications is for cardiac stress tests
performed on over 40 million patients since 1991 (Bristol-Myers Squibb 2003). One
brand of99mTc, Cardiolite, used in stress tests, generated sales of $304 million from
January 2007 to September 2007, not including physician and additional treatment costs
induced with each scan (Ghose 2008). Sales of99Mo, the precursor to 99mTc, are projected
3
-
7/31/2019 99mTc Production in North America
4/34
by the Society of Nuclear Medicine (SNM) to increase 15% over the next decade (Atcher
2008) with Bio-Tech Systems market research predicting 7.5-9.4 % growth per year
between 2006-2012, or 50-70% increase during that time (Committee 2009). Patent
expiration and the entry of generics into the market are sure to have some effect on these
projections.
Diagnosing
Brain
Disorders >57,000 brainscans per year
69,000projected in
2010
Treating Thyroid
Cancer>608,000 thyroid
imaging and therapyper year 652,000projected in 2010
Imaging
Lungs for
Blood Clots>2,070,000lung scansper year2,350,000
projected in2010
Diagnosis and
Monitoring of
Cancer>818,200 cancer
imagesper year
2,053,000 projectedin 2010
Scanning
Bones for
Infection>2,825,000bone scans
per year3,255,000
projected in2010
Diagnosing
Coronary Artery
Disease>8,092,500
cardiologyscans per year
13,407,500projected in 2010
Figure 1: Sample Medical Uses of99m
Tc (Atcher 2008: data Bio-Tech Systems, images Journal of
Nuclear Medicine)
The absorbed radiation dose to specific organs of reference man (70 kg) is estimated
by the manufacturer and is provided as a package insert with the formulation kit. Since
99mTc is a gamma emitter the radiation dose is generally whole body although due to
pharmakinetics localized organs will concentrate 99Tc at differing rates. Estimated
absorbed doses (D) range from 1.8 mGy (breast) to 55.5 mGy (upper large intestine wall)
due to the intravenous injection of 1110 MBq (30 mCi) of Cardiolite used for a stress
4
-
7/31/2019 99mTc Production in North America
5/34
test (Bristol-Myers Squibb 2003). For reference, the whole body limit for the public is 1
mGy/y and to a radiation worker is 50 mGy/y. Both limits are in addition to that received
from background and as a result of medical procedures.
Tc-99m is an ideal radiopharmaceutical useful to medical applications because of
an effective (inside the body) half-life (T1/2~ 6 h) long enough to complete a study with
adequate concentrations remaining within the organ of interest, but also short enough to
keep patient doses to a minimum. The short half life insures that a patient undergoing an
outpatient procedure has cleared the99m
Tc from all organs, by decay or excretion,
approximately six hours from injection (Bristol-Myers Squibb 2003). The primary photon
emitted, at 140 keV with 89% yield, is energetic enough to escape the body without
significant attenuation (ICRP38 1999). The photon is detected by a gamma camera in
which scintillators (plastic or NaI-Tl crystals depending upon brand) convert the energy
of the photon into flashes of light which are converted by photomultiplier tubes into
electrical pulses and are counted electronically.
TcTcMo 99(gamma)IC)(orIT99mMeV1.2Edecaybeta99 max =
T1/2:99Mo 67 hr
99mTc 6 hr
99Tc >105 yr (essentially stable)
Figure 2: Decay of99
Mo
Due to its short half-life99m
Tc cannot be produced directly since it will decay
away in shipping before reaching the patient. Therefore it is eluted with sterile saline
from a resin column containing its parent isotope 99Mo. Mo-99 is currently generated for
use in the US as a fission product of uranium-235 (235U). The remainder of this paper will
5
-
7/31/2019 99mTc Production in North America
6/34
discuss the status and challenges of producing99
Mo on an adequate scale with a focus on
the emerging and proposed process changes.
Current Supply Processing
Abbreviation Name235
U Content
NU natural uranium 0.7%
LEU low enricheduranium
20%
HEU high enricheduranium
> 20%
HEU weapons grade 85%Table 1: Uranium Enrichment Grades
The current system of producing 99Mo utilizes ~ 93% 235U as a requirement of the
approved drug application on file with the Food & Drug Administration (FDA).
Enrichment to 93% is not performed commercially in the US as HEU is generally used
for nuclear weapon production. As a result the HEU used for 99Mo production is shipped
from the Y-12 National Security Complex (Y-12) in Oak Ridge TN as it is removed from
defense stockpiles (Mangusi 2000). HEU is then shipped to Chalk River Laboratories
(CRL) in Canada to be manufactured into targets. Published Nuclear Regulatory
Commission (NRC) records indicate shipments of HEU from the US to Canada for99Mo
production totaling from 10-25 kg per year (NRC).
At NRU MDS Nordion manufactures HEU targets, formed as pins of uranium -
aluminum alloy within an aluminum cladding. These targets are then inserted via
irradiation ports into the National Research Universal (NRU) reactor at CRL. The NRU
reactor operates with neutron fluence (th) of approximately 21014 to 31014 n cm-2s-1.
Up to 20 targets may be irradiated at any one time and can remain within the reactor for
6
-
7/31/2019 99mTc Production in North America
7/34
five to seven days. The99
Mo is generated as a fission product of235
U and will occur in
about 6% of all fissions (Reed 1953). Targets are monitored to determine removal at
optimum times. The time of removal is determined based on the buildup of99
Mo from the
fission of235U. Due to an equilibrium, additional irradiation is not productive as the 99Mo
will be lost due to decay as it is generated and therefore approximately 97% of the 235U
remaining in the target will become waste (Committee 2009). The targets are removed
after irradiation, allowed to cool in water for up to half a day, then are transferred to an
associated hot cell facility (so named for its ability to handle the great heat and intense
radiation of the targets) in shielded casks. Processing then must proceed quickly to
minimize 99Mo losses due to radioactive decay. Approximately 1% of the generated 99Mo
is lost to decay per hour after irradiation. For reference, the Cintichem reactor (which
produced 99Mo in the US until 1989 when the reactor was shut down) typically yielded
600 Ci of99
Mo per target irradiated (Vandegrift 2007). Although processing equipment
must be housed within a heavily shielded facility, the equipment itself is actually bench-
scale and has a footprint similar to that of a large dining room table (Committee 2009).
Ingrowth of Mo-99 vs Thermal Neutron Fission of
U-235
Mo-99
U-235
0 5 10 15 20 25 30
t (days)
Activity
Figure 3: Buildup of99
Mo in Reactor
7
-
7/31/2019 99mTc Production in North America
8/34
Figure 4: Babcock & Wilcox hot cell facility (B&W)
At the hot cell facility the cladding is punctured and gaseous fission products are
removed, such as133
Xe and131
I, which are also valuable medical isotopes. Hot nitric acid
dissolves the target assembly, forming a nitrate solution containing uranium,
molybdenum and other fission products. This solution is then poured through an alumina
column (Al2O3) that adsorbs the nitrates. The column is washed with additional nitric
acid to elute excess uranium and other fission products but which leaves the molybdenum
bound within the column matrix. The addition of sodium hydroxide will elute the
purified99
Mo (Saha 1998). This process typically yields recovery greater than 85-90%
(Committee 2009). The removed 99Mo is shipped immediately to Mallinckrodt (dba
Covidien) in Maryland Heights MO, or Bristol-Myers Squibb Medical Imaging (dba
Lantheus Medical Imaging) inNorth Billerica, MA, where they then manufacture the
generator. Manufacture includes adsorption of ammonium molybdate on Dowex-1
anion exchange resin and washing with concentrated HCl removes any other impurities.
The 99Mo is pH adjusted to form ammonium molybdate ((NH4)6Mo7O24) and is eluted
from the column with dilute HCl (Saha 1998). The alumina column utilizes positively
charged beads which adsorb the 99Mo as 99MoO42- (Zolle 2006). The column is
8
-
7/31/2019 99mTc Production in North America
9/34
autoclaved for sterility. Additional purification and testing per pharmaceutical regulations
is performed before the generator is packaged. The column has a higher affinity for the
99Mo than the
99mTc ensuring that the
99mTc is preferentially eluted from the column.
Elution of99Mo, or breakthrough, does occur. The amount of breakthrough is a quality
control parameter of the manufacturer (Radioactivity 2009). The NRC regulates the
amount of99
Mo that may be given to humans to no more than 0.15 kBq99
Mo per 1 MBq
of99mTc (0.15 Ci 99Mo per 1 mCi 99mTc) and that, upon receipt of the generator, the
concentration of99
Mo in the first elution is verified to conform to the limit (Permissible
2007).
Due to the short half-life (~66 hour) of99Mo this process must take place quickly.
Most nuclear medicine departments rely on continuous, multiple shipments of
99Mo/99mTc generators per week to meet treatment demands. This supply chain is
currently capable of supplying99
Mo /99m
Tc from reactor extraction to patient in less than
48 hours (assuming no delays from any individual step). Industry practice currently sells
bulk99Mo as six day curies which is nominally defined as the activity of the 99Mo in
the generator six days after leaving the producer. Current weekly demand (2nd quarter of
2008) is estimated at 6000 six-day-curies. This weekly demand equates to about 40,000
Ci of99Mo at end-of-irradiation (out-of-reactor) (Robertson 2008).
At the hospital or clinic a radiopharmacist will calculate the amount of99mTc that the
decaying 99Mo has generated based on the transient equilibrium that exists between the
two species. The 99mTc is eluted from the column with sterile saline (0.9% NaCl), as
sodium pertechnetate (Na99m
TcO4), and then is mixed with other reagents, according to
9
-
7/31/2019 99mTc Production in North America
10/34
commercial kits like Cardiolite, into final patient dosage forms. The column is re-used,
or milked, until all the 99Mo has decayed and the column is exhausted.
Figure 5: Elution of Generator Column (generalized) (Sampson 1994)
Figure 6: Decay-growth Relationship in a99
Mo-99m
Tc Generator (Saha 1998)
10
-
7/31/2019 99mTc Production in North America
11/34
Figure 7: General Stages of99Mo Production from Irradiation to Patient. (Making 2008)
Proposed changes to current supply processes
The most pressing change to the development of a 99Mo source in the US is switching
the targets from HEU to LEU, but other options are being considered as well. Other
means of producing 99Mo include photo fission of235U, neutron activation of98Mo, and
photo neutron interactions of100Mo . Construction of new reactors or conversion of
existing reactors or the possible use of accelerators to meet increasing demand is also a
focus of study as well as the use of alternate radionuclides or imaging protocols.
11
-
7/31/2019 99mTc Production in North America
12/34
Using LEU Instead of HEU
Converting existing reactors to LEU targets would significantly reduce the security
concerns that surround the current supply process. The Schumer Amendment to the
Energy Policy Act of 1992 restricts the export of HEU to research and test reactors, of
which NRU is one, for any reason unless they can prove that they cannot use LEU as
target material (Schumer 1992). Currently, NRU can only utilize targets manufactured
using HEU due to the availability of waste storage capability and, mostly, FDA drug
approvals, but research is progressing towards conversion to LEU targets. Clearly the
change from HEU to LEU targets would seem to be simple, but there are technical and
regulatory hurdles to overcome.
Substituting LEU directly in place of HEU targets is possible. However, it would
reduce the 99Mo yield to approximately 20% of that generated from the HEU. This is due
to the reduced number of235
U atoms available to fission (table 1). The lower yield can be
overcome by irradiating five times more targets in the reactor. A reactor designated for
99Mo production could be designed to have the space available to irradiate many more
targets than current reactors can hold due to other user demands on them. Unfortunately,
irradiating five times more targets would result in five times the volume of separation
wastes which could quickly become a storage and disposal problem. Using LEU targets
would also increase the generation of fissile 239Pu due to neutron capture by the higher
proportion of238U. Generating 239Pu, a fissile and long-lived isotope, is also a national
security concern. However plutonium is generally insoluble, and can therefore be
separated from liquid wastes and ultimately be used to make mixed-oxide (MOX) fuel for
nuclear power plants (see NRC Fact Sheet: Mixed Oxide Fuel).
12
-
7/31/2019 99mTc Production in North America
13/34
Figure 8: Production of239
Pu Resulting from Neutron Capture by238
U
The increase in waste volumes is an important issue. If current target dissolution
processes are maintained the waste storage facilities may not be able to accommodate a
five-fold increase. Waste facility modification is one possibility for NRU however its
also possible that in making certain changes they may not need to make that investment.
It is possible that storage tanks are currently operating below capacity and by converting
liquid wastes to solid form, waste storage will not be impacted (Committee 2009).
Instead of irradiating more targets, the targets themselves could be altered to contain
more LEU (increase the target volume). This approach would solve the issue of
throughput in the reactor but could be limited by space restrictions in the reactor.
Changing the composition of the target itself to contain more 235U by changing the
density of the LEU target could also compensate for the reduced yield of99Mo. Current
targets incorporate a uranium-aluminum alloy, with uranium density of approximately 1.6
g/cm3. Converting to uranium metal could yield a higher density of uranium of about 8
g/cm3. The change in density would approximate the mass of235U given the difference in
enrichment (Committee 2009). Argonne National Laboratory (ANL) has been testing a
LEU metal foil target within aluminum, nickel or zinc fission barriers that are
encapsulated within a cylinder of aluminum cladding. Using foils, these targets, test
irradiated in Indonesia, have been compatible with both acidic and alkaline dissolution
processes (Vandegrift 1997). Commercial production using LEU targets is undergoing
testing at the University of Missouri Research Reactor (MURR) which operates a
maximum flux of 6 1014 n cm-2 sec-1(Committee 2009).
13
-
7/31/2019 99mTc Production in North America
14/34
The main difficulty in utilizing LEU foils has been the method to manufacture the
foils themselves. The LEU foils have been manufactured using three differing methods,
by casting: pouring molten metal into molds, hot-rolling: heating the metal above its
recrystallization temperature and then rolling it out to form sheets, or cold-rolling: rolling
the metal out at room temperature to maintain its crystal structure. The casting method
has produced irregular targets that, although useable, would make qualifying the
uniformity of the targets more difficult and possibly more expensive to produce. Hot- and
cold-rolling are both very labor intensive and therefore expensive to produce but result in
better target uniformity (Committee 2009).
After irradiation, a target of LEU or HEU must be chemically processed utilizing one
of only two processes that are FDA approved: MDS Nordion or Cintichem. The
(proprietary) MDS Nordion acid-dissolution process is in use for the isotope generated
from NRU. The Cintichem process was purchased by the Department of Energy (DOE)
from Cintichem Inc. upon the decommissioning of their Tuxedo NY reactor and is the
focus of proposed modifications to convert to an LEU target for US supplies. According
to available sources both processes are quite similar to each other, however a specific
description of the MDS Nordion process is unpublished (Committee 2009).
Ensuring that changes to chemical processing suitably achieve the removal of
contaminating fission products is the focus of chemistry process research for LEU
targets. The prime contaminants are isotopes of iodine (I), rhodium (Rh) and silver (Ag).
Researchers from Argonne National Lab (ANL), and the Universities of Illinois and
Texas have performed tests focusing mostly on the dissolution of the LEU foil targets in
nitric acid as opposed to a mixture of nitric acid and sulfuric acid to avoid the generation
14
-
7/31/2019 99mTc Production in North America
15/34
of sulfate species that can be corrosive in some storage situations (Gause 1982).
Irradiation and processing of foils for the generation of99Mo has been tested in multiple
countries with cooperation and funding from the International Atomic Energy Agency
(IAEA). Testing of LEU foils and a modified Cintichem process with acid target
dissolution are US, Indonesia, Poland, Chile, Korea, Libya, Pakistan, India and Romania
(Goldman 2007). Results confirmed that utilizing LEU foils and a modified acid
dissolution process are feasible alternatives to the current processes. Argentina has been
producing99
Mo utilizing LEU foils dissolved under a modified Cintichem process with
alkaline conditions for their smaller scale domestic imaging needs. The Argentinean
process was able to ultimately recover 92% of the available 99Mo but less than 10% of the
contaminating131
I. The131
I is chemically reduced and therefore adsorbed onto the
alumina column in addition to the 99Mo instead of being removed during purification
(Vandegrift 2007). The differences in contamination purification efficiency that exist,
dependant upon acid or alkaline dissolution, will likely determine the process ultimately
chosen if LEU foils are used to generate US 99Mo supplies.
It is not a requirement that a solid form of LEU is irradiated to generate usable 99Mo.
A possible source of domestic99
Mo is the Medical Isotope Production System (MIPS)
plan proposed by Babcock & Wilcox (B&W). They have proposed to build an aqueous
homogeneous reactor (AHR), also known as a solution reactor, which will utilize a LEU
fuel salt dissolved in acid and water. Various solution forms are possible: uranyl nitrate
(UO2(NO3)2), uranyl sulfate (UO2SO4), or uranyl fluoride (UO2F2). In this reactor the
uranium in the solution is both reactor fuel and target material. This system would
eliminate the necessity to manufacture or encapsulate a target and un-fissioned uranium
15
-
7/31/2019 99mTc Production in North America
16/34
can be recycled back into the reactor instead of being removed as waste. The extraction
process itself would not differ markedly from the existing adsorption to an alumina
column that has already been in use. B&W has estimated that a reactor could generate
approximately 1100 six-day Ci assuming a 26 hour processing time. They also have
estimated a total waste volume of less than 5 m3 each year consisting generally of
alumina column wastes (Reynolds 2007).
Photo-fission of238
U
An accelerator can be utilized to generate99
Mo by the photo-fission of the238
U
present in natural uranium. In this method a high intensity beam of photons is generated
and directed at 238U (Freeman 2008). This results in nearly the same 6% fission yield for
99Mo, however the cross section for
238U is about 1000 times smaller than the neutron
capture cross section for the fission of235U ( 600 barns for thermal neutron capture,
37 barns for the production of99
Mo assuming 6% fission yield) therefore a very large
photon source would be required (Committee 2009).
Using Molybdenum Instead of Uranium
Eliminating uranium from the process is also a possibility by using other isotopes of
molybdenum and converting them to99
Mo. Four reactions are possible:98
Mo(n,)99
Mo
(neutron activation) occurring inside a reactor, 100Mo(,n)99Mo (neutron emission),
100Mo(p,pn)99Mo and 100Mo(p,2n)99mTc which require an accelerator.
Neutron Activation of98Mo
The activation of enriched 98Mo (natural molybdenum ~ 24% 98Mo) is currently in
use by small producers in Kazakhstan and Romania. The reaction,98
Mo(n,)99
Mo
(neutron activation), requires thermal (~ 0.025 eV) or epithermal (0.025 1.0 eV)
16
-
7/31/2019 99mTc Production in North America
17/34
neutrons. Although this method does generate the desired99
Mo, not all the98
Mo is
activated- analogous to the un-fissioned 235U from the MDS Nordion irradiations. This
presents a problem in the purification to pharmaceutical specifications. Mo-99 produced
in this way is not carrier-free, as it contains 98Mo that behaves chemically identical and
acts as a contaminant. Thus, the 99Mo produced is termed low specific activity and would
require larger generators, and by extension larger shielding around the generator, to
produce the same activity required for a medical procedure. The FDA also limits the
amount of99
Mo that can breakthrough during the final dosage form elution process. The
potential exists for increased elution of undesirable
98
Mo when formulating patient
dosages which, beyond potential danger to a patient, limits the useful lifetime of the
larger generator as it will need to be discarded due to breakthrough levels. As an aside,
this method does not produce the secondary fission products that are of medical use,
namely 131I and133
Xe (Committee 2009).
Ci/g106.3Bq/Ci107.3
Bq/g1034.1
)102.3*98(1.0)102.4*99(9.0
104.17SA:Mo10%withMo
Ci/g108.4Bq/Ci107.3
Bq/g1076.1
102.4*99
104.17SA:free-carrierMo
8
10
3
195
239899
5
10
16
5
2399
=
=
+
=
=
=
=
ss
s
Figure 9: Specific Activity Effect of98Mo Contamination
Accelerator Reactions of100
Mo
The accelerator produced methods explored do have merit. However all three would
require extensive investment in accelerators that can generate very high intensity beams
of protons (500 A) or electrons, to create photons, with sufficient energy to overcome
the significantly smaller cross sections involved in these reactions. It has been calculated
17
-
7/31/2019 99mTc Production in North America
18/34
that the100
Mo(p,pn)99
Mo reaction would require about 100 cyclotrons, assuming
currently used cyclotron production values (Committee 2009).
The final method,100
Mo(p,2n)99m
Tc, results in the direct production of99m
Tc which
has the significant disadvantage of time. Since the 99mTc has such a short half life (~6
hours) any need to ship the generator over distances would reduce the usefulness to the
end-user and, by extension, the patient.
Other Challenges
The current system of shipping HEU from Y-12 in Tennessee to NRU in Canada, and
shipping 99Mo back to the US occurs with many points of contention and supplies are in
almost constant jeopardy. Security concerning fissile materials that could be stolen and
used as an improvised nuclear device (IND) or dirty bomb is a current fear concerning
the use of HEU worldwide. Domestically, pharmaceutical companies must appeal to the
Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC) in order to
secure defense stockpiled HEU from Y-12. Shipping itself is carried out under military
supervision with permission from the Department of Transportation (DOT). All three
agencies must grant approval for the raw material (HEU) to be shipped. Delays have
occurred in the past due to reluctance of the various agencies to ship HEU to a private
enterprise out of direct US control (Mangusi 2000). Legally, the Schumer Amendment to
the Energy Policy Act of 1992 states that HEU cannot be exported from the US with very
few exceptions allowed one of them being no approved alternative to the use of HEU
such as that for generation of99Mo (10CFR110). As a result of these confounding factors,
18
-
7/31/2019 99mTc Production in North America
19/34
periodic challenges to NRU and Cintichem (when in operation) regarding the transport
HEU have arisen (Tilyou 1989).
Another serious hurdle to overcome in the development of domestic production
capability or changes to current processes is the approval for human use by the Food &
Drug Administration (FDA). The production of any pharmaceutical, including medical
radioisotopes, can only be performed according to a documented series of steps submitted
and held by the FDA as a Drug Master File (DMF). Any changes to any step in the
process will require an amendment to the DMF and subsequent approval by the FDA.
The initial approval, such as would be required for a new manufacturer like B&W or
MURR, includes a New Drug Application (NDA) which outlines all parameters for the
manufacture of the drug (Brown 2005). Before any product can be given to patients, three
complete production runs must be performed and evaluated for safety, efficacy and purity
according to current Good Manufacturing Practices (cGMP) (21CFR210, 21CFR211).
Even changes to an existing NDA or DMF requires an evaluation to ensure changes do
not deleteriously affect drug safety, efficacy and purity. The product is tested against
internal protocols but must conform to accepted pharmacopeial standards such as those
covered in the United States Pharmacopeia (USP), the European Pharmacopeia (EP),
Japanese Pharmacopeia (JP) or others. Some standard testing parameters are pH,
concentration, radiochemical and radionuclide purity (Sodium 2009).
Although extensive testing prior to process conversion should establish that 99Mo
generated will perform no differently than the current process, the FDA will review all
submitted data to determine the actual performance. This qualifying requirement adds
considerable costs beyond infrastructure. These include protocol and system development
19
-
7/31/2019 99mTc Production in North America
20/34
for the three qualification runs, raw materials testing, production costs of the three
complete and typical-sized production runs, materials to complete the assembly of
generators, final testing of the completed generators for all three runs and the costs
associated with analyzing, compiling and completing the NDA. Each manufacturer will
have to independently navigate this process which has been estimated to be at least
$250,000 (Brown 2005). An FDA presenter to the National Academy of Science (NAS)
suggested review lead times ranging from 4-18 months before approval although they
could be extended significantly if FDA determined that human clinical trials would be
required (Committee 2009).
Figure 10: FDA Regulatory Process (adapted from Brown 2005)
Another major stumbling block to domestic production is the capital investment that
would be required. Estimates of costs and timelines vary extensively since no new reactor
facilities or isotope separation facilities have begun construction in the US for over two
decades.
20
-
7/31/2019 99mTc Production in North America
21/34
Reactor NameConstruction
Start
Criticality
Date
Univ. Texas-Austin 1986 1992Watts Bar 1 1973 1996
Comanche Peak 1 1974 1990Comanche Peak 2 1974 1993Seabrook 1 1976 1990Limerick 1 1974 1990
Table 2: Recent US Reactors
A facility dedicated to isotope production utilizing solid or liquid targets would need
to construct a reactor in addition to hot cell and waste stream facilities. Commissioning a
new reactor for isotope production is a lengthy process. Internationally, start of
construction to commission of a reactor has been 6-8 years based on the ETRR2 in Egypt,
FRM II in Germany and OPAL in Australia. An existing reactor, such as MURR, would
need to create the capability to chemically separate the99
Mo in an associated hot cell.
Construction and commissioning of a new processing facility could take 3-5 years and
has been estimated to cost $30 million to $40 million by the director of MURR, Ralph
Butler (Committee 2009). Time estimates given are minimums and delays are always a
possibility due to regulatory demands or design/construction changes (Committee 2009,
Reynolds 2007).
An example of the challenges to a new reactor facility is the Canadian MAPLE
(Multipurpose Applied Physics Lattice Experiment) experience. Two reactors, MAPLE I
and II, were planned and constructed to supply the worldwide 99Mo demand from one
reactor and to utilize the other as a backup and research reactor. Costs projected to
complete the MAPLE reactors (to replace the aging NRU) were estimated to be $130
million. During the testing phase MAPLE I was discovered to posses a positive
21
-
7/31/2019 99mTc Production in North America
22/34
temperature coefficient of reactivity. The temperature coefficient of reactivity, or
multiplication factor, describes the number of neutrons available per generation per
degree change in coolant temperature. A positive value indicates that there are more
neutrons available as each fission generation progresses and the power of the reactor will
increase as the temperature increases. A negative value indicates that there are fewer
neutrons available for each subsequent generation and the power of the reactor will
decrease as the temperature increases. The reactors were designed to have a temperature
coefficient of reactivity of -0.12 mMW-1 but testing measured +0.28 mMW-1
. Where
is the effective multiplication factor and is dependant upon the absorption cross section
of the fuel and moderator, the enrichment of the fuel (the MAPLE reactors were designed
to utilize HEU fuel), the arrangement of the fuel rods and the coolant capabilities. This
means that the inherent safety requirements of the reactor design were not met.
Additional investigation could not determine the cause of the positive temperature
coefficient of reactivity and in mid-2008 the project was cancelled (Magnus 2008).
22
-
7/31/2019 99mTc Production in North America
23/34
The Four Factor Formula: = f p
Symbol Name Meaning Dependence
Reproduction
Factor (eta)
The number of fission neutrons produced per absorption in
the fuel.
fuel
enrichment
fThe thermalutilizationfactor
Probability that a neutron that gets absorbed does so in thefuel material.
compositionand geometryof fuel(reactordesign)
pThe resonanceescapeprobability
Fraction of fission neutrons that manage to slow downfrom fission to thermal energies without being absorbed.
moderator/fuelratio (reactordesign)
The fast fissionfactor
cross sectionratio,moderator/fuelratio, fuelgeometry(reactordesign)
Themultiplication
factor
-
Table 3: The Four Factor Formula - Temperature Coefficient of Reactivity
Babcock & Wilcox (B&W) assumes that five years will be required from conceptual
design (which has begun) to commercial operation for their AHR. They have formed a
partnership with a pharmaceutical company, Covidien, to assist with production and
processing practice with B&Ws AHR design. An AHR has never been built at the
proposed scale in the US and as such budget projections, based on experience, are
lacking. The directors of MURR have also indicated their intention to commercially
produce 99Mo and have solicited funding for studying the design and construction of
facilities (Scully 2009).
23
-
7/31/2019 99mTc Production in North America
24/34
A recurring theme with regard to developing any definite plans for the construction of
domestic isotope production is securing capital from companies or federal budgets.
Obtaining adequate funding is a major barrier to production. Government and private
enterprise, including the medical community, seem to realize the importance of supplying
99Mo. Unfortunately, funding for such projects will likely be hampered by the current
economic crises and investments of such huge amounts of money may not be deemed
appropriate or truly urgent until some recovery occurs. Waiting for funding will,
obviously, put the ideal timelines out of grasp extending the time until the US can depend
on its own isotope supply.
Other Treatment Options
There are other treatment options available to practitioners when supplies of
99Mo/
99mTc are limited. Since cardiac and bone scanning account for ~75% of
99mTc
applications the examination of alternate treatment modalities will focus on replacements
in those areas. The main competitors to 99mTc are thallium-201 (201Tl), positron emission
tomography (PET) utilizing rubidium-82 (82Rb), nitrogen-13 (13N) ammonia, or fluorine-
18 (18
F) as tracers and computed tomography (CT).
The largest competitor to 99mTc for cardiac perfusion studies is 201Tl. Thallium-201
can be used as a replacement for99mTc but they are often utilized in conjunction with
each other. Thallium-201 is produced by irradiating a natural thallium target in the
external beam of the 60 Brookhaven cyclotron with 31 MeV protons (Lebowitz 1975).
The nuclear reaction is203
Tl(p,3n)201
Pb. The201
Pb decays to201
Tl in 9.4 hours. Thallium-
201 is not considered an ideal isotope for imaging because it decays to mercury-201
24
-
7/31/2019 99mTc Production in North America
25/34
(201
Hg) and emits multiple characteristic x-rays resulting in lower quality images. It also
tends to concentrate in the kidneys due to its longer half life (73 hours) resulting in a
higher radiation dose to that organ than may be acceptable (Committee 2009).
Figure 11: Decay and Emissions from Tl-201 (Lombardi 2006)
The tracers used in PET include rubidium-82 (82
Rb), which is cyclotron produced
requiring 500 MeV protons. The nuclear reaction is 85Rb(p,4n)82Sr. The 82Sr (T1/2 ~ 26 d)
decays to82
Rb (T1/2 ~ 75 s) within a82
Sr/82
Rb generator (similar to the99
Mo/99m
Tc
generator). Since the half life of the 82Rb is so short the generator is milked by
continuous infusion directly into the patient. Use of82
Rb with PET is limited by
generator availability and insurance / Medicare reimbursement. Additionally, radiation
doses to medical personnel who administer the tests can be significantly higher, possibly
limiting the number of tests which could be performed at a center (Schleipman 2006).
Ammonia with 13N is also used. This form of nitrogen has a 10 minute half life and
therefore is made within a hospitals cyclotron. Obviously this limits treatment to
geographical regions with cyclotron availability. Fluorine-18 (18
F) is actually superior to
99mTc for bone scanning however it has limited cardiac use since it is currently not a
reimbursable treatment. The use of PET has increased dramatically, however it is
estimated that the bulk of treatments with 99mTc are performed in private cardiology
25
-
7/31/2019 99mTc Production in North America
26/34
-
7/31/2019 99mTc Production in North America
27/34
not impacted negatively from the changes to their treatment plans, it is also likely that
some were. A negative impact would be to health, insurance reimbursement, scheduling
or trust in the abilities of their care providers.
Additional challenges are constantly evolving. For example, in response to the
National Academy of Science report, Medical Isotope Production Without Highly
Enriched Uranium, released in January 2009, Representative Edward J. Markey (D-Mass)
has stated his intent to introduce legislation to pressure change to eliminate HEU in the
generation of99
Mo and also to ensure supply reliability (Congressman 2009). As of this
writing no legislation has been introduced. It is unknown how any additional legislation
would impact the building of a significant source of99Mo in the US. More active and
consistent support for changes to the supply chain comes from the IAEA and the global
nuclear community who financially support and test the experimental changes explored
throughout this paper.
It seems that there is a favoring of converting worldwide production to the utilization
of LEU with only minor consideration given to other production methods - although this
may be more of a political drive rather than a technical necessity. Regardless, more time
and money has already been invested in research to ensure that conversion to LEU will
be successful. Given that LEU appears to be the preferred raw material for a domestic
production facility I am intrigued by the possibilities of the AHR system. This reactor
designs ability to use the uranium in uranyl nitrate both as target material and as reactor
fuel appears economical to me. I also prefer its ability to essentially recycle the un-
fissioned uranium back into the reactor as this could dramatically reduce the amount of
waste that would require further processing or long-term disposal or security. I am
27
-
7/31/2019 99mTc Production in North America
28/34
optimistic that the B&W / Covidien partnership can ultimately succeed in navigating the
NRC licensing and FDA approvals to achieve their goal of producing a majority of the
US demand for99m
Tc.
There are many hurdles to ensuring stable US supplies: security, economy,
infrastructure and the technical changes to processing procedures in moving away from
HEU. Ensuring that the supply can be achieved while ensuring homeland security and
environmental safety is also important. Of ultimate importance however is supplying
99Mo/
99mTc to patients in a prompt and dependable manner because it is important to the
health and longevity of the American population.
28
-
7/31/2019 99mTc Production in North America
29/34
29
Figure12:Q
uickReview
ofProcesses
-
7/31/2019 99mTc Production in North America
30/34
Bibliography
Amersham Health, Medi-Physics, Inc. Ceretec: Kit for the Preparation of TechnetiumTc99m Exametazime Injection. Brochure. Arlington Heights: Author, 2005.
Atcher, R. W., R. W. Brown, and J. P. Norenberg. "Mo-99 Production." An Outlook onNew Sources of Mo-99 and Other Medical Radionuclides. Proc. of 2008International Meeting on Reduced Enrichment for Research and Test Reactors,Washington DC. Oct. 2008. Society of Nuclear Medicine. Jan. 2009.
Babcock & Wilcox Technical Services Group, Inc. "B&W Names R&D Lead forMedical Isotope Initiative." Press release. Lynchburg, VA. 25 Feb. 2009.
Bristol-Myers Squibb Medical Imaging. Cardiolite Kit for the Preparation of TechnetiumTc99m Sestamibi for Injection. 2003.
Bristol-Myers Squibb Medical Imaging. TechneLite Technetium Tc 99m Generator.Brochure. North Billerica: Author, 2005.
Brown, R. W. "Mo-99 Production." The Radiopharmaceutical Industry's Effort toMigrate Toward Mo-99 Production Utilizing LEU. Proc. of 2005 InternationalMeeting on Reduced Enrichment for Research and Test Reactors, Boston MA.Nov. 2005. Council on Radiopharmaceuticals and Radionuclides (CORAR). Jan.2009 .
Brown, Roy W. "Drug Master File Development and FDA Filings for LEU-ProducedMedical Radionuclides." NAS Committee on the Production of Medical IsotopesWithout HEU. Keck Center, Washington DC. 11 June 2007.
Bushberg, Jerrold T., John M. Boone, and Edwin M. Leidholdt. The Essential Physics ofMedical Imaging. 2nd ed. Philadelphia: Lippincott Williams & Wilkins, 2002.
Chandra, Ramesh. Nuclear Medicine Physics: The Basics. 5th ed. Philadelphia:Lippincott Williams & Wilkins, 1998.
Committee on Medical Isotope Production Without Highly Enriched Uranium. MedicalIsotope Production Without Highly Enriched Uranium. Rep. 14 Jan. 2009.National Academy of Sciences. 14 Jan. 2009.
"Congressman Edward Markey - Jan. 14, 2009 - Report: Nuclear Bomb Materials Can BeEliminated From Production Of Medical Supplies." Congressman EdwardMarkey - Home. 14 Jan. 2009. 6 Feb. 2009.
Conner, C. "Mo-99 Production." Progress in Developing Processes for Converting Mo-99Production from High- to Low-Enriched Uranium - 1998. Proc. of 1998
International Meeting on Reduced Enrichment for Research and Test Reactors,Sao Paulo Brazil. Oct. 1998. Chemical Technology Division, Argonne NationalLaboratory. Jan. 2009 .
Covidien. "B&W and Covidien to Develop U.S. Source of Key Medical Isotope." Pressrelease. Mansfield, MA. 26 Jan. 2009.
CRP on Production of Mo-99 from LEU or Neutron Activation. International AtomicEnergy Agency. 26 Feb. 2009.
30
-
7/31/2019 99mTc Production in North America
31/34
Diamond, Bill. "Yields of Mo-99 from Alternate Production Methods." Oct. 2008.TRIUMF: Canada's National Laboratory for Particle and Nuclear Physics. Nov.2008 .
Dunn Lee, Janice, comp. Proposed License to Export HEU to Canada for Use in the NRUReactor to Produce Medical Radioisotopes. Rockville: Nuclear Regulatory
Commission, 2001.Federal Register, Pg 48921 Vol. 61, No. 181 (1996).Freeman, Tami. "Medical Isotope Supplies: A Game Plan for the Future." Editorial.
Medical Physics Web. 8 Dec. 2008. 19 Feb. 2009.
Gause, E. P., L. G. Stang, D. R. Dougherty, E. Veakis, and J. Smalley. Characterizationof Radioactive Large Quantity Waste Package of the Union Carbide Corporation.Rep. no. BNL-NUREG-30247R. Upton NY: Brookhaven National Laboratory,1982.
Ghose, Carrie. "Generic Drug Could Mean More Money for Cardinal Health." ColumbusBusiness First 21 Mar. 2008. American City Business Journals Inc. Jan. 2009
.Goldman, Ira N. "Mo-99 Production." Fostering New Sources of Mo-99 for InternationalNuclear Medicine Needs - The Contribution of the IAEA Coordinated ResearchProject on Molybdenum-99 Production from LEU or Neutron Activation. Proc. of2008 International Meeting on Reduced Enrichment for Research and TestReactors, Washington DC. Oct. 2008. International Atomic Energy Agency. Jan.2009.
Goldman, Ira N., Natesan Ramamoorthy, and Pablo Adelfang. "Mo-99 Production."Progress and Status of the IAEA Coordinated Research Project: Production ofMo-99 Using LEU Fission or Neutron Activation. Proc. of 2007 InternationalMeeting on Reduced Enrichment for Research and Test Reactors, Prague, CzechRepublic. Sept. 2007. International Atomic Energy Agency. 18 Feb. 2009.
Homogeneous Aqueous Solution Nuclear Reactors for the Production of Mo-99 andOther Short Lived Radioisotopes. Tech. no. TECDOC-1601. Vienna, Austria:IAEA, 2008.
Hot Cell Facility. Babcock & Wilcox Company. B&W Technical Services Group, Inc.Hot Cell Facility. 25 Mar. 2009.
International Commission on Radiation Protection: Radiation Dose to Patients fromRadiopharmaceuticals. Publication no. 80. 3rd ed. Vol. 28. Annals of the ICRP,1999.
International Commission on Radiation Protection: Technetium-labeled MIBI Tc-99m.Publication no. 80. 3rd ed. Vol. 38. Annals of the ICRP, 1999.
31
-
7/31/2019 99mTc Production in North America
32/34
Jarousse, C., P. Colomb, M. Febvre, N. Morcos, and P. Anderson. CERCA Mo-99Annular Cans Target Manufacturing Development for ANSTO. Tech. 22 Mar.2004. European Nuclear Society. 16 Dec. 2008.
Lebowitz, E., M. W. Greene, R. Fairchild, P. R. Bradley-Moore, H. L. Atkins, A. N.Ansari, P. Richards, and E. Belgrave. "Thallium-201 For Medical Use." Journalof Nuclear Medicine 16 (1975): 151-55.
Lombardi, Max H. Radiation Safety in Nuclear Medicine, Second Edition. 2nd ed. CRC,2006.
Lyman, Edwin. "Using Bilateral Mechanisms to Strengthen Physical ProtectionWorldwide." Editorial. Nuclear Weapons & Global Security. 26 Aug. 2008.Union of Concerned Scientists. 26 Feb. 2009.
Magnus, Ben. "Over Budget, Overdue and, Perhaps, Overdesigned." Canadian Medical
Association Journal 178 (2008): 813-14. Canadian Medical Association Journal.20 Mar. 2008. Canadian Medical Association. 12 Jan. 2009.
Making Medical Isotopes: Report of the Task Force on Alternatives for Medical-IsotopeProduction. Rep. Vancouver, BC, Canada: TRIUMF, 2008.
Mangusi, John. "Application to USNRC for License to Export Nuclear Material(10CFR110)." Letter to Ronald D Hauber. 23 Oct. 2000. Nuclear RegulatoryCommission Reading Room. 2001. United States Nuclear RegulatoryCommission. 15 Jan. 2009 .
MDS Nordion. Sr-82 Fact Sheet. Brochure. Ottawa, ON, Canada: Author, 2008.MDS Nordion. Tl-201 Fact Sheet. Brochure. Ottawa, ON, Canada: Author, 2008."Mo-99/Tc-99m Generators: From Reactor to Patient." Monthly Scan 11 (July 2006): 1-
2. Radiopharmacy, Inc. July 2006. 16 Dec. 2008.
Mushtaq, A., Massod Iqbal, Ishtiaq H. Bokhari, Tariq Mahmood, Tayyab Mahmood,Zahoor Ahmad, and Qamar Zaman. "Neutronic and Thermal Hydraulic Analysisfor Production of Fission Molybdenum-99 at Pakistan Research Reactor-1."Annals of Nuclear Energy 35 (2008): 345-52.
"NRC: Electronic Reading Room." NRC: Home Page. United States Nuclear RegulatoryCommission. .
Permissible Molybdenum-99, Strontium-82, and Strontium-85 Concentrations, 10 CFR 35.204 (2007).
Preliminary Draft Report of the SNM Isotope Availability Task Group. Rep. June 2008.Society of Nuclear Medicine. Dec. 2008 .
"Radioactivity." United States Pharmacopeia. Vol. 31. Rockville: United StatesPharmacopeial Convention, 2009. 340.
Reed, George W., and Anthony Turkevich. "Uranium-235 Thermal Neutron FissionYields." Physical Review 92 (1953): 1473-481.
32
-
7/31/2019 99mTc Production in North America
33/34
Reynolds, W. Evans. "Babcock & Wilcox Medical Isotope Production System." MedicalIsotope Production. United States Nuclear Regulatory Commission, RockvilleMD. Dec. 2007.
Reynolds, W. Evans. "Babcock & Wilcox Medical Isotope Production System Status."Medical Isotope Production System Status Update. United States Nuclear
Regulatory Commission, Rockville MD. Oct. 2008.Robertson, David. "University of Missouri and MU Research Reactor Center." DOEIsotope Production & Applications Workshop on Nation's Need for Isotopes:Present and Future. Rockville MD. Aug. 2008. Isotope Production andApplications. Dept. of Energy Office of Science for Nuclear Physics and Office ofNuclear Energy. Dec. 2008 .
Saha, Gopal B. Fundamentals of Nuclear Pharmacy. 4th ed. New York: Springer, 1998.Sampson, Charles B. Textbook of Radiopharmacy: Theory and Practice. 2nd ed. Taylor
& Francis, 1994.Schleipman, A. Robert, Frank P. Castronovo, Marcelo F. Di Carli, and Sharmila Dorbala.
"Occupational Radiation Dose Associated With Rb-82 Myocardial PerfusionPositron Emission Tomography Imaging." Journal of Nuclear Cardiology 13(2006): 378-84.
""Schumer Amendment" to the Energy Policy Act of 1992." Reduced Enrichment forResearch and Test Reactors (RERTR) [Nonproliferation]. 25 Jan. 2009.
Scully, Sarah. "MU Research Reactor poised for global production of medicalradioisotope -." Columbia Missourian. 9 Jan. 2009. 25 Mar. 2009.
Society of Nuclear Medicine. Isotope Shortage Survey Final Results. Raw data.Http://interactive.snm.org/docs/Isotope Survey Results 11-3-08.pdf. 3 Nov. 2008.
"Sodium Pertechnetate Tc 99m Injection." United States Pharmacopeia. Vol. 31.Rockville: United States Pharmacopeial Convention, 2009. 3336.
Stabin, Michael G. "Effective half life of technetium-99m tests." Health Physics Society.10 Oct. 2003. 10 Mar. 2009.
10 CFR 110.42(a)(9), 134 (1992).21 CFR 21021 CFR 211Tilyou, Sarah M. "DOE's Suspension of Enriched Uranium Shipments at Savannah River
Puts Mo-99/Tc-99m Supply in Jeopardy." The Journal of Nuclear Medicine 30(1989): 1139+.
United States. Department of Public Health. Food and Drug Administration. Guidance forIndustry: Changes to an Approved Application for Specified Biotechnology andSpecified Synthetic Biological Products. Rockville: Center for Drug Evaluationand Research, 1997.
United States. Environmental Protection Agency. Environmental Impact Statement forProposed Medical Isotope Production. 129th ed. Vol. 60. Washington DC:Federal Register, 1995.
33
-
7/31/2019 99mTc Production in North America
34/34