Shielding material selection for the disposal of radioactive waste from 𝟗𝟗𝐌𝐨 production

Abstract

This study focuses on the radiation shielding performance of 4 types of concrete for the storage, road transport and final disposal of cemented intermediate-level liquid (radioactive) waste generated during the production of fission-based Mo-99, i.e. (𝑛,𝑓99Mo, in concrete drums. There is a large and growing worldwide demand for the medical radionuclide 99mTc. However, this isotope has a short half-life of only ~6 hours, so that the delay between its production and arrival at remote users, will cause most of the activity to be lost through radioactive decay. However, 99mTc has a longer-lived precursor, 99Mo with a half-life that is 11 times longer than that of 99mTc, so that it is far more practical to produce 99Mo and then deployed it in a 99Mo-99mTc generator from which the short-lived medical radionuclide 99mTc is eluted at nuclear medicine clinics. When a fission-based process is used to produce 99Mo, the product is almost carrier-free; such high specific activity (HSA) 99Mo is, internationally, in high demand as a longer-lived precursor for 99mTc. The production of (𝑛,𝑓)99Mo is by the irradiation of fissile target plates in a research reactor. HALEU (High-Assay Low Enrichment Uranium) target plates are irradiated for as long as 200 hours. After discharge and initial cooling, the target plates are dissolved via alkaline digestion. Iodine and molybdenum are chemically separated from the solute. The combined activity yield of the 2 medically useful and marketable fission products -- 99Mo and 131I -- from the fission of the uranium isotope 235U, is in the order of 10 %, i.e. the other 90 % of the fission products constitute radioactive waste. The waste from the chemical processing of irradiated fissile target plates in an isotope production nuclear reactor, naturally separates into 3 streams: (1) high-level solid uranic residue, (2) intermediate-level liquid waste (ILLW) and (3) gaseous radionuclides. The high activity ILLW is first held in retention tanks for circa 90 days, and are then mixed with cement powder and vermiculite and agitated. The resulting cementitious slurry is transferred into 100-litre stainless-steel drums placed inside large concrete drums. The cementation process solidifies and immobilises the liquid radioactive waste. The concrete drums are placed in interim on-site storage; as soon as radiological acceptance criteria are met, these drums are transported by road to the national radioactive waste disposal site, for disposal in near-surface trenches. This study presents a literature review of radioactive waste management, focusing on the immobilisation of ILLW via cementation and the combination of an engineered barrier system and a natural barrier system for disposal. The physics theory of the attenuation of ionising photons is presented and a simple but robust recipe is derived.

The dimensions of the concrete drums are fixed, but there is freedom for optimising the concrete material in a quest for lower transmitted dose rates, as well as the possibility of loading higher initial activities of radioactive waste per drum. This research project therefore quantifies the relative abilities of 4 types of concrete, as shielding material for the large concrete drums in which the cemented liquid radioactive waste is disposed. The 4 types of concrete options that are investigated for their abilities to reduce transmitted dose-rates, are (1) ordinary concrete (the baseline case), (2) magnetite concrete, (3) hematite concrete and (4) barite concrete. Design-basis values for (1) the fissile isotope content and uranium vector in target plates, (2) fission power per target plate assembly and (3) irradiation time in the core of an isotope production reactor, were defined. To evaluate the radiation shielding performance of the concrete types, radionuclide activities and the resulting photon emission source terms must first be quantified. Codes in the SCALE 6.2.4 code system along with its nuclear data, were used to quantify the source term in the ILLW for a number of discrete times after the end of the irradiation of the target plates in the reactor. Problemspecific, self-shielded 1-group ORIGEN cross-sections were prepared with a long run of the SCALE sequence, T6-DEPL. These cross-section sets were prepared to allow subsequent fast analyses with the SCALE code ORIGEN. A SCALE ORIGEN calculation model was used to model the irradiation, cooling, chemical partition of the elements during chemical processing, which separates the liquid waste from the solid residue. This ORIGEN calculation model was also used to follow the isotopic evolution of the source term in the intermediate-level liquid radioactive waste in discrete time steps up to 10 years. Calculation models of the 4 concrete drum types, each containing source terms representing different cooling times, were developed for the Monte Carlo radiation transport code MCNP6.2. The contact dose rates on the outer surface of the drums, as well as the Transport Index (TI) dose rate at a distance of 1 m from the drums, were quantified at different decay times of the cemented ILLW. Dose-rate fields were quantified around a single drum as well as at a perimeter fence around a rectangular array of drums were calculated. The shielding performance, i.e. the ability to reduce the transmitted dose rate, of all 4 concretes were measured against the reference case of no shielding. The dose rate reduction performance of the 3 high-density shielding concretes, are also expressed relative to the performance of ordinary concrete. A notable finding was that a shielding wall has to be constructed around the decay yard; such a design was developed. Another investigation concerned the required burial depth of the drums in the trenches. A conservative calculation model was developed and used to arrive at a safe burial depth for the drums.

Conclusions are that barite concrete has the best dose rate reduction performance, but that hematite concrete may be the preferred material choice, based on its complete local supply chain as well as operational experience with constructing durable hematite concrete drums that can survive long road journeys without cracking. Both these high-density concretes offer significant dose rate reduction advantages over ordinary concrete.