Radioactive Waste Issues Related to Production of Fission-based <sup>99</sup>Mo by using Low Enriched Uranium (LEU)
Radioactive Waste Issues Related to Production of Fission-based 99Mo by using Low Enriched Uranium (LEU)
Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT). 2015. Jun, 13(2): 155-161
Copyright © 2015, The Korean Radioactive Waste Society
This is an Open-Access article distributed under the terms of the Creative Commons Attribution Non-Commercial License ( which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original work is properly cited
  • Received : December 05, 2014
  • Accepted : May 11, 2015
  • Published : June 30, 2015
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Muhmood ul, Hassan
Ho Jin, Ryu

Technetium-99m ( 99m Tc) is an important, short-lived decay product of molybdenum-99 ( 99 Mo), and it is considered the backbone of the modern nuclear diagnostic procedures. Since fission of 235 U is the main source of production of 99 Mo, either highly-enriched uranium (HEU) targets or low-enriched uranium (LEU) targets are irradiated in the research reactors. The use of LEU targets is being promoted by the international community to avoid the proliferation issues linked with the use of HEU. In order to define the waste management strategy at the planning stage of establishment of an LEU based 99 Mo production facility, the impact of the use of LEU targets on the radioactive waste stream of the 99 Mo production facility was analyzed. Because the volume of uranium waste is estimated to increase six times, the use of high uranium density targets and the utilization of hot isostatic pressing were recommended to reduce the increased waste volume from the use of LEU based targets.
1. Introduction
Molybdenum-99 ( 99 Mo) is an important radioisotope with a half-life of approximately 66 h. It decays to technetium-99m ( 99m Tc) which has a life of approximately 6 h [1] and it is this radioisotope that forms the backbone of modern nuclear diagnostic procedures. The decay scheme of 99 Mo to 99m Tc is shown in Fig. 1 [2] .
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Decay scheme of 99Mo and 99mTc.
In order to produce fission-based 99 Mo from research reactors, two types of targets have been developed and irradiated, such as highly-enriched uranium (HEU) targets with 235 U enrichments more than 90wt% of 235 U and low-enriched uranium (LEU) targets with 235 U enrichments less than 20wt% of 235 U [3] . It is worth noting that medium-enriched uranium i.e., 36wt% of 235 U, as being used in South Africa, is also regarded as non-LEU from a nuclear security point of view. The production yield of 99 Mo from these aforementioned targets is not comparable. LEU targets have a 99 Mo yield of only 20% of the HEU targets. However, the increased use of HEU has proliferation issues and can be a potential security risk. Therefore, international nuclear security policy is promoting the use of LEU targets in order to minimize the civilian use of HEU [4] .
In order to fulfill the increasing demand of 99 Mo by using LEU targets, approximately five times more LEU tar-gets need to be irradiated than HEU targets which will ultimately increase the volume of radioactive waste. The use of LEU targets may increase the quantity of TRUs (Pu) in the uranium containing solid waste and will make the handling of those more complex [5] .
In this study, we discuss and compare mainly the radio-active waste generated by alkaline digestion of both HEU and LEU targets in order to better assist in the planning for proper handling and disposal of generated waste.
2. Types of the Radioactive Waste
The radioactive waste generated from fission-based 99 Mo production can be due to: target fabrication, assembling of target, irradiation in reactor and processing of irradiated targets [6] . During the fission of 235 U in a reactor, a large number of radionuclides with different chemical and physical properties is formed [7] . After irradiation, the ir-radiated targets are dissolved to extract the 99 Mo. This dissolution can either be acidic or alkaline, but for convenient storage of solid waste and easy separation of fission gases, alkaline dissolution is preferable. A representative equation of alkaline dissolution is given below [8] :
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where the value x, ranging from 0.1 < x < 0.16, is the fraction of the aluminum dissolved under nitrite formation. Dissolution and purification processes remain the main stream of all three types of radwaste (gaseous, liquid and solid). The produced solid, liquid, and gaseous waste may be a combination of low-level waste (LLW) and intermediate-level waste (ILW). However, no high-level waste (HLW) is produced from 99 Mo production facilities as it does not contain long-lived or alpha-emitting radionuclides with activity > 400 Bq/g and it does not produce heat > 2 kW/m 3 [9 - 13] . It is envisioned that large amounts of radioactive waste will be produced from 99 Mo production facilities using LEU targets for a large production of 99 Mo. Handling and treatment of the generated waste is dependent on its form and activity. In the case of a large production facility, a waste storage facility should be constructed in order to limit the radiation exposure of the workers and the environment.
- 2.1 Gaseous Waste
99 Mo production process consists of the following main steps ( Fig. 2 ) [14] :
  • i. Target fabrication
  • ii. Irradiation of targets
  • iii. Dissolution of irradiated targets
  • iv. Separation and purification of final product
  • v. Packaging and supply to the end users (e.g. medical centers)
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99 Mo production flow diagram (HEU & LEU) [14].
Dissolution of targets remains an important step from the point of view of radioactive waste generation. Irradiated 235 U targets are loaded into a hot cell where, after separation from aluminum and other claddings, 235 U plate or foil is inserted into a dissolver to separate the 99 Mo from other fission products.
During this process, a number of volatile and radioactive gases are produced. The main active gases and their activity produced in the dissolver for LEU targets containing 16 g uranium, 235 U contents = 3.2 g and irradiated for a time of 12 h in neutron flux of 1.33×10 14 n cm -2 s -1 along with their half-lives are given in Table 1 [6] .
Major off-gases produced during dissolution in curie[6]
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Major off-gases produced during dissolution in curie [6]
Activity and volume of the produced gases depends on the cooling time of the targets. In current practices, the offgases are stored in especially designed/shielded delay/decay cascades and are released into the environment after decay as required by the applicable regulations. The off-gases handling system of the facility should be sufficient to hold certain volumes of these fission gases. The amount of the fission gases to be released into the environment is based on the regulatory limits set by the licensing authority/regulatory body (e.g., in NUREG-0472, the normal release of 133 Xe at the boundary of the facility dose rate limits should not exceed 500 mrem/year (5 mSv/year) for the total body and 3000 mrem/year (30 mSv/year) for skin). Stringent monitoring arrangements are required in order to avoid any leakage of these gases into the processing area or environment.
Production of radio xenon during the dissolution process is of major concern as radio xenon is a signature gas of any nuclear test, and its release/leakage from the isotope production facility may generate spurious signals for international monitoring systems (IMS) as can be seen in Fig. 3 . In order to control this situation CTBTO has recommended the release limit for radio xenon as less than 5 GBq/day from radioisotope facilities. Different studies have also been carried out and found that a maximum release of 5 GBq/day can be a safe limit and has no effect on IMS.
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Mixing of xenon emissions from some medical isotope production facilities (circle) with the nuclear test signatures (cross) on the “weapons” side of the Kalinowski Discrimination Line (dashed line) [15].
Moreover, production of hydrogen during alkaline digestion of target plates can be a safety concern. The accumulation/leakage of hydrogen may cause an explosion and may be a source of release of fission product gases into the process area.
- 2.2 Liquid Waste
In both alkaline and acidic dissolution processes, liquids are being used in the dissolver in order to dissolve the irradiated targets. In the production of 99 Mo, the main source of liquid waste arises from the dissolver and filtra-tion process. The resulting liquid waste may possibly be a combination of both LLW & ILW. For technical reasons, these two types of waste should be kept separate. Inappropriate design of the storage tanks may cause leakage of the liquid waste. The storage container material should have chemical compatibility with the liquid waste to avoid corrosion and degradation accidents.
Liquid waste generated in Mallinckrodt, Netherlands(93% enriched HEU UAlxplate targets, 3000 Ci)[16].
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Liquid waste generated in Mallinckrodt, Netherlands(93% enriched HEU UAlx plate targets, 3000 Ci) [16].
The generation of liquid waste is different for both types of fuel targets used for fission based 99 Mo production. In order to produce the same amount of 99 Mo from the LEU target, it must contain uranium at ≥5 times the uranium in the HEU target. As a result, three to six times more dissolver solution will be required [7] . The liquid waste will contain almost all the unburned uranium and 239 Pu. According to IAEA [7] , about 26 times more 239 Pu is produced in LEU targets than HEU targets with comparable 99 Mo yield which increases the disposal cost significantly.
- 2.3 Solid Waste
The main source of solid waste in the 99 Mo production process consists of filtered uranium and TRU, ion exchange resins, and absorber columns. It may also consist of charcoal filters, different hot cell filters, valves, pumps, tubes, etc. The target design can also be a major contributor (e.g., activated aluminum cladding and its cut pieces removed from an annular target etc.). The generated solid waste needs to be categorized according to activity and its heat contents. Moreover, it needs to be segregated as low-level, medium-level and alpha-bearing waste for proper treatment and storage/disposal.
3. Discussion
The processing of LEU targets for 99 Mo production comparable to HEU targets potentially sees an increase in the radioactive waste as listed in Table 3 [17] . The pro-duced waste may include increased uranium waste, spent dissolver solution, wipes, and TRU waste. It is reported that 180 g of uranium containing waste is produced by dissolving LEU targets with 19.75% enrichment of 235 U and containing total uranium of 93.7 g to produce 545 Ci of 99 Mo [17] . This amount of uranium waste is six times the comparable production of 99 Mo from HEU targets.
The radioactive waste generation from HEU and LEU targets for99Mo production[17]
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The radioactive waste generation from HEU and LEU targets for 99Mo production [17]
During the separation of 99 Mo from the dissolver so lution a larger filter or column bed is required as large amounts of dissolver solution is needed to be filtered in the case of the use of LEU targets. Also the presence of 26 times more 239 Pu in the waste stream of LEU target process ing may have a significant impact on the waste classification as well as on its disposal and management cost [17] . Although the production yield of 239 Pu is very low, the alpha activity of 99 Mo should be inspected carefully because alpha contamination of 99 Mo should be minimized for the radiation safety of patients. Generally, the alpha contamination limit of 10 -7 μCi/mCi of 99 Mo is recommended [18] .
Volume reduction of process waste can be achieved by the development of high-density targets, the optimized dissolution and extraction processes, and the consolidation of waste products. If high density targets with a uranium density higher than 8 g-U/cm 3 are used, the production yield of 99 Mo can be enhanced more than three times theoretically, because current LEU targets have a uranium density of 2.6 g-U/cm 3 . In order to increase the uranium density of LEU targets, high-density uranium alloys should be used as dispersion fuel particles for the plate-type targets. New, high-density targets with enhanced production yield which are compatible with current digestion technology have been under investigation [19 , 20] . It is also proven that the selection of such advanced target designs should have less safety issues during the irradiation and the dissolution processes. Moreover, the use of volume reduction technologies may be considered and applied to treat and reduce the volume of produced waste for interim storage or final disposal. For example, Synroc technology can be applied to reduce the waste volume and immobilize the long-lived intermediate level waste by using hot isostatic pressing [21] . For example, ANSTO Australia is carrying out detailed engineering of a Synroc plant for 5000 ℓ/year intermediate level waste from ANSTO 99 Mo production facility by utilizing hot isostatic pressing technology of ceramic waste form [21] .
In order to ensure the safety in waste management, the above-mentioned issues of increased radioactive waste production should be considered at the time of planning and establishment of a 99 Mo production facility. Although an increase in the volume of process waste for LEU-based production of 99 Mo has been reviewed generally in this study, waste management procedures, guides, clearance levels, interim and permanent storage, and transportation arrangements need to be addressed quantitatively before the commissioning and operation of such a facility.
4. Conclusions
Radioactive waste management has key importance during the planning, development, and commissioning phase of a molybdenum production facility. With the use of the LEU targets in a 99 Mo production facility, significant increase in liquid and solid waste is expected. Therefore, it is necessary to consider safety issues related to the radioactive waste management in the production of 99 Mo via LEU targets, with the development and establishment of appropriate waste handling, storage, and treatment arrangements on or off-site of the 99 Mo production.
This study was supported partly by the National Nuclear R & D Program of the Ministry of Science, ICT and Future Planning (NRF No. NRF-2014M2C1A1029177).
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