Below are the first 10 and last 10 pages of uncorrected machine-read text (when available) of this chapter, followed by the top 30 algorithmically extracted key phrases from the chapter as a whole.
Intended to provide our own search engines and external engines with highly rich, chapter-representative searchable text on the opening pages of each chapter. Because it is UNCORRECTED material, please consider the following text as a useful but insufficient proxy for the authoritative book pages.
Do not use for reproduction, copying, pasting, or reading; exclusively for search engines.
OCR for page 184
Medical Isotope Production without Highly Enriched Uranium Appendix D Alternative Molybdenum-99 Production Processes There are two primary processes for producing molybdenum-99 (Mo-99): fission of uranium-235 (U-235) and neutron capture of molybdenum-98 (Mo-98). These are shown schematically in Figures D.1 and D.2, respectively. The fission of U-235 produces a large number of fission products, including Mo-99. The mass distribution of these fission products is shown in Figure 2.5. The rate of production, which is of interest here, is proportional to several conditions as illustrated in the equation below: where R = rate of reaction (i.e., number of reactions per unit time and volume), which is related to the amount of the new substance that can be produced n = the number of target nuclei present (i.e., the target nuclei density in atoms per unit volume) = the flux of particles causing the reaction (neutrons per cm2 per second) σ = the probability that the reaction will occur, expressed as an area To understand whether a particular method is better than another these parameters must be considered as illustrated in the following comparisons.
OCR for page 185
Medical Isotope Production without Highly Enriched Uranium FIGURE D.1 Schematic representation of the uranium-235 fission process. N = neutrons and FPs = fission products. FIGURE D.2 Production of Mo-99 from neutron capture. N = neutron. The most commonly used alternative method for producing Mo-99 involves the neutron capture on an enriched target of Mo-98 (natural occurrence of Mo-98 is 24.13 percent), which is illustrated schematically in Figure D.2. The fission cross section for thermal fission of U-235 is approximately 600 barns1 which represents a very high probability. Of this, approximately 1 1 barn = 1 × 10−24 cm2.
OCR for page 186
Medical Isotope Production without Highly Enriched Uranium 6.1 percent results in the production of Mo-99 or about 37 barns. The production cross section for the 98Mo(n,γ)99Mo reaction is about 0.13 barn for thermal neutrons, a factor of almost 300 less than the fission process even accounting for the 6.1 percent fission yield for Mo-99. There are 6 stable isotopes (92, 93, 94, 95, 96, 97) of Mo and two very long-lived isotopes (98 is >1012 years and 100 is >1018 years). Both Mo-98 and Mo-100 have long enough half-lives that they exist in nature and can be used as target material. Thus the ability to produce large amounts of Mo-99 from the direct reaction route would depend upon the availability of a high flux reactor that could compensate for the lower cross section. For example, typical fluxes from the National Research Universal (NRU) reactor are around 1.5 × 1014 neutrons per cm2 per second while the High Flux Isotope Reactor (HFIR) at Oak Ridge has a flux of 1015 neutrons per cm2 per second, more than enough to be competitive in producing large amounts of Mo-99 via the (n,γ) approach.2 However, these additional neutrons are not free and would add to the costs of producing Mo-99 by this method. However, the Mo-99 produced by this process has a very low specific activity3,4 because most of the Mo in the product is Mo-98. The specific activity for fission-produced Mo-99 is two to four orders of magnitude higher than from the neutron capture process (Ottinger and Collins, 1996). This has practical implications for using neutron capture Mo-99 in medical isotope procedures: First, the technetium generators that are used for fission-produced Mo-99 would have to be redesigned to use neutron capture-produced Mo-99. A larger technetium generator column would be needed, which would increase the size of the generator and the size and weight of its shield. A larger volume of liquid would be required to elute Tc-99m from the column, which would require all of the current Tc-99m kits (e.g., see Table 2.1) to be reformulated. In addition, the useful lifetime of the generator would be reduced due to the potential for higher breakthrough5 of the Mo-99. This would require users to purchase additional generators. 2 If desired, the isotope could also be enriched in Mo-98 using mass separation processes. 3 Specific activity is defined as the amount of radioactivity per unit mass as is usually expressed in terms of Becquerel’s per gram or curies per gram. 4 Delft University researchers are examining the feasibility of using Szilard Chalmers reactions to increase specific activities. However, the yields from this process are likely to be small, and a great deal of development work would likely be required to get to a useful, practical process, if indeed it is possible at all. See http://www.tudelft.nl/live/pagina.jsp?id=29b23a65-485b-44ee-9210-f460e363c2c6&lang=en. Accessed October 23, 2008. 5 When the generator is eluted to obtain Tc-99m a very small amount of Mo-99 is released. The generator can no longer be used when the amount of Mo-99 in the eluted solution exceeds a certain level. The amount of breakthrough is roughly proportional to the amount of molybdenum present, both radioactive Mo-99 and nonradioactive Mo-98.
OCR for page 187
Medical Isotope Production without Highly Enriched Uranium TABLE D.1 Comparison of Fission and Neutron Produced 99Mo 235U(n,f)99Mo 98Mo(n,γ)99Mo Produces high specific activity 99Mo Produces low specific activity Mo-99 Requires enriched 235U target Requires highly enriched Mo-98 target Complex chemical processing Simple chemical processing Requires dedicated processing facility Requires high flux neutron source Generates high-level radioactive waste Generates minimal waste SOURCE: Modified from S. Mirzadeh, Oak Ridge National Laboratory. Table D.1 compares the two methods of production. Another point to consider, although of secondary importance, is the fact that several other radionuclides of medical importance are coproduced in the fission process and would require an alternative source (in particular 131I and 133Xe) in the case of a neutron capture process. To make use of the neutron capture approach a number of technical challenges must be overcome not the least of which is the availability of the desired Tc-99m in a useful chemical form and of the same quality as the fission product for use with the many radiopharmaceutical kits now on the market. This point applies for all of the alternative processes discussed below. ACCELERATOR PRODUCTION There have been a number of proposals for accelerator production of Mo-99 as well as for direct production of Tc-99m. One accelerator-based approach essentially mimics the reactor production route in that the accelerator becomes the source of neutrons, which are then used to produce fission in a blanket of U-235 surrounding the neutron source. The required fluxes would be difficult to achieve in the required geometry to be competitive with reactor-generated neutrons. Such an accelerator would be expensive to build and operate although less expensive than a new reactor. Another approach would be to use an electron beam to generate high-intensity photons which in turn would be used to initiate a nuclear reaction on enriched Mo such that 100Mo(γ,n)99Mo creates the desired product (TRIUMF, 2008). The same issues as discussed above holds for this approach in addition to the technical challenges associated with producing a high-energy electron machine with sufficient beam flux to be able to produce sufficient Mo-99 to be competitive. That said, there are discussions around the design of electron linacs capable of accelerating tens of milliamps of electrons. For both of these accelerator approaches multiple machines would be required since the fluxes of neutrons and photons would not be sufficiently
OCR for page 188
Medical Isotope Production without Highly Enriched Uranium high to be competitive with a reactor. The cost of construction and operation of multiple machines would have to be analyzed to determine if a business case could be made for these approaches. Another approach is photo-fission of U-238 using natural or depleted uranium targets. The challenge is the same as is mentioned for the other photon induced reaction (100Mo(γ,n)99Mo); that is, the need for a very high intensity beam to overcome the factor of about 1000 smaller cross section for this reaction versus neutron fission of U-235, although the fission yields are almost identical (approximately 6 percent). The other option that has been explored is the direct production of Tc-99m from the 100Mo(p,2n)99mTc. The biggest disadvantage with this approach is that the final product (the one used in nuclear medicine procedures) is directly produced and has a short half-life (6 hours). Thus, its usefulness would be greatly hampered if it needed to be shipped great distances to the end users. Even a network of suppliers would face a challenge. Takács et al. (2002) report that the cross section for the direct production of Tc-99m from enriched Mo-99 would be approximately 17 mCi/μAh. At this level even a very high beam current facility (500μA protons) and irradiation periods of a day (i.e., 24 hours), the most that could be produced in a single facility would be < 200 Ci per day. To meet the needs of the United States there would have to be more than 25 cyclotrons dedicated to this process. This does not take into account the losses associated with transport and chemical efficiencies for separating the Tc-99m from the target matrix. A single site might be able to become self-sufficient but this would not help the larger community. Takács et al. (2002, 2003) explored the production of Mo-99 from the 100Mo(p,pn)99Mo reaction. Their results indicated a thick target yield (40–45 MeV) of 3.8 mCi/μAh. The daily production for a similar cyclotron would be about 50 Ci thus about 100 cyclotrons would be required for this approach. The other approach would be through the spallation (high-energy projectile collides with the target nucleus with enough energy that a very large array of products is produced) of a target to produce Mo-99. The production rate of Mo-99 from most reasonable target materials would be at best many orders of magnitude lower than the reactor methods and two orders of magnitude lower than the above accelerator reactions and thus not a viable approach. From this analysis there are few viable alternative approaches to the supply of Mo-99 or Tc-99m for widespread distribution. With the termination of the Maple reactor project, alternative approaches need to be explored in comparison to the cost of constructing and commissioning a new reactor facility, especially with photon-induced fission with U-238.
OCR for page 189
Medical Isotope Production without Highly Enriched Uranium REFERENCES Ottinger, C. L., and E. D. Collins. 1996. Assessment of Potential ORNL Contributions to Supply of Molybdenum-99. Oak Ridge National Laboratory Report No. ORNL/TM-13184. Oak Ridge, TN: ORNL. Takács, S., F. Tárkányi, M. Sonck, and A. Hermanne. 2002. Investigation of the natMo(p,x)96mgTc nuclear reaction to monitor proton beams: New measurements and consequences on the earlier reported data. Nucl Instrum Methods Phy Res B 198:183-196. Takács, S., Z. Szűcs, F. Tárkányi, A. Hermanne, and M. Sonck. 2003. Evaluation of proton induced reactions on 100Mo: New cross sections for production of 99mTc and 99Mo. Radioanal Nucl Chem 257:195-210. TRIUMF. 2008. Making Medical Isotopes: Report of the Task Force on Alternatives for Medical-Isotope Production. Available at http://admin.triumf.ca/facility/5yp/comm/Report-vPREPUB.pdf.