1.Introduction

Molybdenum-99 (⁹⁹Mo) is an important radioisotope with a half-life of approximately 66 h. It decays to technetium- 99m (^99mTc) 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 ⁹⁹Mo to ^99mTc is shown in Fig. 1[2].

In order to produce fission-based ⁹⁹Mo from research reactors, two types of targets have been developed and irradiated, such as highly-enriched uranium (HEU) targets with ²³⁵U enrichments more than 90wt% of ²³⁵U and lowenriched uranium (LEU) targets with ²³⁵U enrichments less than 20wt% of ²³⁵U [3]. It is worth noting that mediumenriched uranium i.e., 36wt% of ²³⁵U, as being used in South Africa, is also regarded as non-LEU from a nuclear security point of view. The production yield of ⁹⁹Mo from these aforementioned targets is not comparable. LEU targets have a ⁹⁹Mo 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 ⁹⁹Mo by using LEU targets, approximately five times more LEU targets 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 radioactive 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 ⁹⁹Mo production can be due to: target fabrication, assembling of target, irradiation in reactor and processing of irradiated targets [6]. During the fission of ²³⁵U in a reactor, a large number of radionuclides with different chemical and physical properties is formed [7]. After irradiation, the irradiated targets are dissolved to extract the ⁹⁹Mo. 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]:

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 ⁹⁹Mo 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/m³ [9-13]. It is envisioned that large amounts of radioactive waste will be produced from ⁹⁹Mo production facilities using LEU targets for a large production of ⁹⁹Mo. 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.

2.1Gaseous Waste

⁹⁹Mo production process consists of the following main steps (Fig. 2) [14]:

Dissolution of targets remains an important step from the point of view of radioactive waste generation. Irradiated ²³⁵U targets are loaded into a hot cell where, after separation from aluminum and other claddings, ²³⁵U plate or foil is inserted into a dissolver to separate the ⁹⁹Mo 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, ²³⁵U contents = 3.2 g and irradiated for a time of 12 h in neutron flux of 1.33×10¹⁴ n cm^-2s^-1 along with their half-lives are given in Table 1[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 ¹³³Xe 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.

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 ⁹⁹Mo, the main source of liquid waste arises from the dissolver and filtration 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.

The generation of liquid waste is different for both types of fuel targets used for fission based ⁹⁹Mo production. In order to produce the same amount of ⁹⁹Mo 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 ²³⁹Pu. According to IAEA [7], about 26 times more ²³⁹Pu is produced in LEU targets than HEU targets with comparable ⁹⁹Mo yield which increases the disposal cost significantly.

2.3.Solid Waste

The main source of solid waste in the ⁹⁹Mo 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.Table 2

3.Discussion

The processing of LEU targets for ⁹⁹Mo production comparable to HEU targets potentially sees an increase in the radioactive waste as listed in Table 3 [17]. The produced 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 ²³⁵U and containing total uranium of 93.7 g to produce 545 Ci of ⁹⁹Mo [17]. This amount of uranium waste is six times the comparable production of ⁹⁹Mo from HEU targets.

During the separation of ⁹⁹Mo from the dissolver solution 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 ²³⁹Pu in the waste stream of LEU target processing may have a significant impact on the waste classification as well as on its disposal and management cost [17]. Although the production yield of ²³⁹Pu is very low, the alpha activity of ⁹⁹Mo should be inspected carefully because alpha contamination of ⁹⁹Mo should be minimized for the radiation safety of patients. Generally, the alpha contamination limit of 10^-7μCi/mCi of ⁹⁹Mo 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/cm³ are used, the production yield of ⁹⁹Mo can be enhanced more than three times theoretically, because current LEU targets have a uranium density of 2.6 g-U/cm³. 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, highdensity 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 ⁹⁹Mo 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 ⁹⁹Mo production facility. Although an increase in the volume of process waste for LEU-based production of ⁹⁹Mo 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 ⁹⁹Mo 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 ⁹⁹Mo via LEU targets, with the development and establishment of appropriate waste handling, storage, and treatment arrangements on or off-site of the ⁹⁹Mo production.