A.G. Armstrong, “Development of Nuclear Safeguards Approaches for a Molten Salt Reactor Design”, M.S. Thesis, Nuclear Engineering, Texas A&M University, College Station, TX (2021).
The original molten salt reactor (MSR) concepts were developed at Oak Ridge National Laboratory (ORNL) in the 1960s. Increasing energy demands and concerns about global CO2 emissions have emphasized the need for advanced nuclear energy systems, such as MSRs. The inclusion of MSRs in the Generation IV reactor designs bolstered global interest in the development of MSR technologies beyond the Nuclear WeaponsrnStates (NWSs). This has created proliferation concerns relating to MSRs, especially as they can be considered small modular reactors (SMRs), which makes them an economically attractive entry into nuclear reactor development for Non-Nuclear Weapons States (NNWS) and potentially for States that are looking to acquire nuclear material for illicit weapons programs. However, MSRs present unique challenges for the applicationrnof nuclear safeguards, and there are currently no existing technical solutions. One potential technology is hybrid K-edge densitometry (HKED), which could be used to determine isotopic concentrations at strategic points in the system to support implementation of material balance areas (MBAs) that are required to determine if any special nuclear material (SNM) is unaccounted for. The value of material-unaccounted-for (MUF) must be lower than one significant quantity (8 kg for plutonium, 75 kg of 235U for low enriched uranium, 25 kg of 235U for high enriched uranium) in a specified time frame, known as a material balance period (MBP). This regulation is to ensure that the MSR can meet international safeguards requirements as applied by the International Atomic Energy Agency (IAEA). Therefore, to simulate SNM accountancy measurements results from a reference MSR design, a uranium-based reactor core design was developed as part of this research. The reactor was designed to operate in the thermal neutron spectrum with a power output of 300 MWth. The reactor core consisted of a hexagonal design with 61 fuelsalt channels and 108 blanket-salt channels, each connected to salt plena on the top and bottom of the core. The core was surrounded by a graphite radial reflector and a boronrncarbide (B4C) neutron absorber, which are enclosed in a Hastelloy-N reactor vessel. The fuel-salt was composed of uranium tetrafluoride (UF4) with 235U enrichment of 3.5%, constituting 20% of the fuel-salt by core weight fraction, and mixed with molten lithium and beryllium fluoride (FLiBe) at a temperature of 900 K. The fuel-salt channels were surrounded by depleted UF4 blanket-salt channels. The fuel-salt and blanket-salt both hadrnthe same density (2.892 g/cm3). This proposed reactor model was developed using Monte Carlo N-Particle Transport (MCNP version 6.2) code and the Standardized Computer Analyses for Licensing Evaluation (SCALE version 6.2) code system. These models were used to compare effective neutron multiplication factor values under steady-state operating conditions. Fuel burnup simulations were performed using the MCNP model only. Based on the results of SNM production and fuel-salt flow rates, MSR safeguards approaches were developed. Results of the MCNP fuel burnup simulations were used to determine SNM quantities over time in each of the three designated MBAs for the MSR by utilizing HKED as an SNM assay methodology at the KMPs. A combined (systematic and random)rnmeasurement uncertainty of 0.28% provided for HKED by the IAEA ITV document was assumed. When considering the combined three loops of the MSR, the IAEA conditions (MUF < SQ, MUF < 3σMUF, and 3σMUF < SQ) were met for both 235U and plutonium for the MBP of 12 months. Further analysis was done considering a multi-module site with four MSRs. This analysis showed that the 3σMUF < SQ condition would not be satisfiedrnfor either 235U or plutonium for the 12-month MBP. This work demonstrates that both MCNP and SCALE can be used to model anrnMSR in developing NMA strategies for a molten salt reactor core design. Results of the fuel-salt depletion were used to determine that the proposed MBPs of one month and three months meet the three IAEA safeguards compliance conditions (MUF < SQ, MUF < 3σMUF, and 3σMUF < SQ) for both 235U and plutonium.