99mTc Production in North America

7/31/2019 99mTc Production in North America

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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


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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

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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

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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

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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

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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

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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

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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

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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

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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)

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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.

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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).

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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).

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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

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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

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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)

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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

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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,

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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

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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.

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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

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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).

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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).

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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


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(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

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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

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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.

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29

Figure12:Q

uickReview

ofProcesses


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Documents

99mTc Production in North America