The key industrial method for producing ⁹⁹Mo is production of the radionuclide as one of the ²³⁵U fission fragments. ²³⁵U is irradiated with neutrons in a nuclear reactor (both heterogeneous and homogeneous nuclear reactors can be used) and then processed in radiochemical laboratories, where ⁹⁹Mo is chemically extracted from fission products. Both highly enriched uranium (HEU) and low enriched uranium (LEU) can be used to produce ⁹⁹Mo by the fragmentation method. To date, almost all world producers, with the exception of Russia, are either in the final stages of transferring production from highly enriched uranium to low enriched uranium, or are already producing ⁹⁹Mo using LEU. This is due to the problems of non-proliferation of nuclear materials and the prevention of the likelihood of terrorist threats. A number of experimental studies have been carried out on the basis of the VVR-ts research reactor. Experimental studies included the study of the effect of LEU targets on the reactivity reserve of the VVR-ts reactor, irradiation of these targets in experimental channels and separation of ⁹⁹Mo from them. The paper presents the results of producing and separating ⁹⁹Mo from targets with LEU material. It is shown that it is necessary to improve the processing technology to increase the production of fragmented ⁹⁹Mo from LEU.

fragmentation ⁹⁹Mo, research reactor VVR-ts, highly enriched uranium, low enriched uranium, target for ⁹⁹Mo production

Topical nowadays is to increase the production of the ⁹⁹Мо radionuclide for diagnosis of oncological diseases (World Nuclear Association. Radioisotopes in Medicine). Russia’s major producers of fission ⁹⁹Мо using HEU are the Research Institute of Atomic Reactors (JSC SSC RIAR) in Dimitrovgrad and the L.Ya. Karpov Research Institute of Physical Chemistry (JSC RIPC) in Obninsk covering in full the demand of Russian medicine for this isotope (Molybdenum-99 (2022) Russian Nuclear Community). In connection with threats to nonproliferation of nuclear materials, an issue has arisen regarding the conversion of the fission ⁹⁹Мо production to LEU targets (Review of Potential Molybdenum-99/Techetium-99m Production Technologies 2010; Non-HEU production technologies for molybdenum-99 and technetium-99m 2013; Muranaka 2022; Program for converting civilian nuclear research reactors and molybdenum-99 production targets from HEU to LEU 2012). Achieving this requires undertaking a set of research and technology activities.

HEU-based production of the fission ⁹⁹Mo radionuclide was started at JSC RIPC in 1985. The technology for manufacturing and irradiating targets has been improved continuously leading to a major increase in production of the desired radionuclide. Currently, the potential output amounts to about 400 Ci (commercial grade) per week.

Activities were undertaken in 2015–2019 to justify the geometry and performance of LEU targets, as well as the possibility for manufacturing such targets (Zevyakin et al. 2022). In 2020–2021, a number of LEU targets were manufactured at JSC RIPC and a range of experimental studies were conducted, including the investigation of the LEU target effect on the VVR-ts reactor reactivity margin, irradiation of these targets in experimental channels, and isolation of ⁹⁹Мо from them. Targets with direct-flow cooling were used (Fig. 1).

Figure 1. 

Design of a target with direct-flow cooling for ⁹⁹Мо production.

A target with direct-flow cooling represents a tubular structure with a water feedthrough for improved removal of heat (Pakholik et al. 2021; Zevyakin et al. 2022). There is a mixture of a uranium-containing material and zinc oxide placed between the internal tube and the external tube. Not more than two targets are installed into the channel at a time.

Standard channels of the MAK and MAK-2 loop devices with standard operating parameters (pressure, flow rate, temperature) were used (Pakholik et al. 2021).

The experimental study was conducted in a depoisoned reactor. The reactivity margin (r_measi) was determined from the control rod (CR) in-core position (Kochnov et al. 2022) with the reactor brought to the minimum controlled power level (MCPL) prior to and after the LEU target loading into experimental channel (EC) 8–9. The experimental study consisted of the following stages.

• Bringing the reactor to the MCPL, fixing the CR position and the primary circuit water temperature (T₁, °С), and reducing power to zero. The reactor reactivity margin (r ₁) was calculated based on experimental data.
• Loading LEU targets into EC 8–9.
• Bringing the reactor to the MCPL, fixing the CR position and the primary circuit water temperature (T₁, °С), and reducing power to zero. The reactor reactivity margin (r ₂) was determined based on experimental data.
• Determining the change in the VVR-ts reactor reactivity margin (r) with the LEU targets placed in EC 8–9.

The reactor reactivity margin (r₁) with the reactor brought to the MCPL during stage 1 was calculated from the fixed CR position (integral characteristics of the control rods are used):

r₁ = r_meas1 (1)

The reactor reactivity margin (r₂) with the reactor brought to the MCPL during stage 3 was calculated taking into account the variation in the primary circuit water temperature (DT, °С) relative to stage 1:

r₂ = r_meas2 - Dr_t, (2)

where Dr_t = a_t·DT is the temperature reactivity effect, %; a_t = ± 0.012%/°С is the VVR-ts reactor temperature reactivity coefficient (Kochnov et al. 2022).

The reactivity margin variation was calculated using the following formula:

Dr = r₂ - r₁ (3)

Table 1 presents the measurement and calculation results.

As it can be seen from Table 1, the positive reactivity increase was 0.042% with the LEU target loaded into the experimental channel of the VVR-ts reactor. In a similar experimental study, a reactivity increase of 0.045% was obtained for a standard HEU target (Table 2). These results show that the neutronic performance of the LEU target is comparable with that of the HEU targets in use. The generation of ⁹⁹Мо in LEU targets during irradiation in the reactor is comparable with the generation of the same radionuclide in HEU targets.

As the result of testing ⁹⁹Мо from the LEU-based target, no excessive values of the radionuclide purity (the ratio of gamma impurities to ⁹⁹Мо with its total activity in a range of 240 to 407 GBq as of the manufacturing date) have been recorded, but there is a radionuclide pair of ⁹⁵Zr + ⁹⁵Nb present in the ⁹⁹Мо solution. As the result of the ⁹⁵Zr decay, ⁹⁵Nb is accumulated (T_1/2=35 days). As the result of the ^99mTc decay and the ⁹⁵Nb accumulation, the final spectrum looks as follows (Fig. 2).

Figure 2. 

⁹⁹Мо solution spectrum: 1 – Тс-⁹⁹m peak (140.66 keV); 2 – Ru-103 peak (496.96 keV); 3 – Zr-95 peak (723.68 keV); 4 – Мо-99 peak (739.06 keV); 5 – Zr-95 peak (756.22 keV); 6 – Nb-95 peak (765.30 keV).

A number of GT-4K molybdenum-technetium generators (Technetium-99m generator of the GT-4K type) with an activity of 4 to 19 GBq were loaded with the obtained ⁹⁹Мо. An analysis of eluate has shown a deviation for the ‘chemical impurities’ parameter, namely an excessive content of manganese. The rest of the parameters are as required by the Manufacturer’s Pharmacopoeial Monograph (MPM).

It can be seen from Tables 4, 6 that the yield of the desired ⁹⁹Мо radionuclides in the two experiments (23.30 and 13.30% respectively) differs greatly from the calculated standard processing values (61%). The radiochemical process is likely to be affected by the presence of a large quantity of ²³⁸U, the concentration of the ZnO diluter, and some other factors.

Experimental studies were conducted based on the L.Ya. Karpov Research Institute of Physical Chemistry to investigate the feasibility of ⁹⁹Мо generation and isolation using a LEU target.

Experiments with a LEU-containing powder target have shown the validity of the neutronic, thermo-hydraulic and strength calculations conducted as part of justifying the safety of the LEU target irradiation in experimental channels of the VVR-ts reactor, including two sequential cycles (no coolant overheating, mechanical changes or target integrity loss).

The possibility has been demonstrated for using the existing ⁹⁹Мо isolation technology. The radionuclide purity of the molybdate-sodium solution is generally as required by regulatory documentation. The ^99mTc obtained from GT-4K molybdenum-technetium generators largely meets the MPM eluate requirements.

In connection with the fact that there are no currently data on the coefficients of extraction, separation, etc., the effective technology with the existing parameters cannot be conclusively carried over from HEU targets to LEU targets due to multiple uncertainties.

For addressing technological issues, experimental activities are required to define the conditions for the target dissolution (рН of the medium, dissolution time, acid concentration, etc.), and the extraction conditions (ratio of the phase volumes, etc.), that is, full-scale technological development of the process.

There is a need for considering a new design of the target as such, that is, for switching over to an exclusively uranium barrier-enclosed target.