The utilization of thorium in Generation IV reactors

Effect of oxide impurities on the dissolution behavior of Th4+, Be2+ and U4+ in fluoride salts

Oxides are one of the most important impurities in the fuel salt of molten salt reactors (MSRs), and excessive oxide impurities pose a risk to the safe operation of MSRs. This study focused on investigating the precipitation behavior between Th4+, U4+, and Be2+ with O2- in the 2LiF-BeF2 (FLiBe) eutectic salt system. The results showed that the solubility of UO2 was 5.52 × 10-3 mol kg-1, and the solubility product (Ksp) of UO2 was 6.14 × 10-7 mol3 kg-3 in FLiBe salt at 650 °C. It was also found that the O2- ion would firstly react with U4+ to form UO2, and then the excessive O2- would react with Be2+ to generate BeO in the FLiBe system. Despite conducting the solubility experiment of ThO2 and titration experiment of FLiBe-ThF4, the system failed to achieve the solubility and the Ksp of ThO2. The main reason for this was that O2- preferentially reacted with Be2+ over Th4+ to form precipitates, in other words, Be2+ exerted a protective effect against Th4+. Above all, this work experimentally demonstrated that in the FLiBe system, O2- preferentially combines with U4+ to form a precipitate, followed by Be2+, while Th4+ is relatively inert.

A Monte Carlo study on burnup treatment in sodium-cooled reactor with Th fuel

Sodium Cooled Reactors is one of the Generation-IV plants selected to manage the long-lived minor actinides and to transmute the long-life radioactive elements. This study presents the comparison between two-designed SFR cores with 600 and 800 MWth total heating power. We have analyzed a conceptual core design and nuclear characteristic of SFR. Monte Carlo depletion calculations have been performed to investigate essential characteristics of the SFR core. The core calculations were performed by using the Serpent Monte Carlo code for determining the burnup behavior of the SFR, the power distribution and the effective multiplication factor. The neutronic and burn-up calculations were done by means of Serpent-2 Code with the ENDF-7 cross-sections library. Sodium Cooled Fast Reactor core was taken as the reference core for Th-232 burnup calculations. The results showed that SFR is an important option to deplete the minor actinides as well as for transmutation from Th-232 to U-233.

Numerical design of thorium and uranium fuel samples irradiation in lead environment

The paper shows capabilities of thorium-lead fuel assembly for design of irradiation experiments on ThO2 and natural UO2 fuel samples using radioisotope neutron source. The main purpose of the current analysis was to determine the irradiation environment in the samples, especially: neutron spectrum, power, activity, reaction rates, production of 233Pa and 239Np as well as breeding of 233U and 239Pu. An advanced three-dimensional numerical model for Monte Carlo radiation transport and burnup simulations was developed using the Monte Carlo Continuous Energy Burnup Code (MCB). The versatility of the assembly gives a perfect opportunity to perform many irradiation experiments for R&D on the thorium and uranium fuel cycle in a different material and geometrical environments.