Neutronic Performance of a Compact 10 MWe Nuclear Reactor with Low Enrichment (ThxU1−x)N Fuel

Recent interest in compact nuclear reactors for applications in space or in remote locations drives innovation in nuclear fuel design, especially non-oxide ceramic nuclear fuels. This work details neutronic modeling designed to support the development of a new nuclear fuel concept based on a mixture of thorium and uranium nitride. A Monte Carlo N-Particle Version 6.2 (MCNP-6) model of a compact 10 MWe reactor design which incorporates (ThxU1−x)N fuel is presented. In this context, a “compact” reactor is a completely assembled reactor which may be emptied of coolant and transported by specialized commercial vehicle, deployed by a C130J aircraft, or launched into space. Core geometry, reflector barrels, and the heat exchange zones are designed to support reduction of overall reactor volume of core components while maintaining criticality with a fixed total fuel mass of 4500 kg. Dense mixed nitrides of thorium nitride (ThN) additions in uranium nitride (UN) in 5 wt.% increments between $$0.05 \le x \le 0.5$$ 0.05 ≤ x ≤ 0.5 have been considered for calculation of $$k_{\infty }$$ k ∞ and $$k_{{{\text{effective}}}}$$ k effective . ThN additions in UN results in a slight increase in the magnitude of the temperature coefficient of reactivity, which is negative by design. The isotopic distribution of the principal actinide inventory as a function of burnup, time, and initial fuel composition is presented and discussed within the context of the proliferation risk of this core design.

Lead-Cooled Fast Reactor Annular UN Fuel Design and Development of Performance Analysis Program

A kind of annular uranium nitride (UN) fuel suitable for lead-cooled fast reactor applications has been designed in this study. The design is directly targeting two main issues of UN fuel: severe swelling and thermal decomposition of UN fuel at high temperatures. A performance analysis program based on FORTRAN programming language has been developed for UN fuel in fast reactors. The program contains heat transfer, fuel stress-strain analysis, cladding stress-strain analysis, fission gas release and fuel-cladding mechanical interaction (FCMI) modules, etc. Extensive code verification has been performed by comparing simulation results obtained with the code and those obtained via the COMSOL Multiphysics platform. Preliminary code validation has been conducted as well by comparing code simulation results with experimental data. The results showed that this program could predict the fuel temperature, stress-strain, and displacement of UN fuel during reactor operation with a reasonable accuracy.

Thermal hydraulic analysis of a PWR loaded with annular fuel rods

In 2006, the final report of the MIT Center for Advanced Nuclear Energy Center the project entitled High Performance Fuel Design for Next Generation PWR’s presented the proposal of an internal and external cooled ring fuel with the objective of increasing the power density of a PWR reactor without compromising the safety margins of the installation. The thermal hydraulic conditions were calculated with the aid of the VIPRE subchannel code, which is a widely used tool in the analysis of nuclear reactor cores. STHIRP-1 is a subchannel code that has been developed at the Departamento de Engenharia Nuclear /UFMG. In order to evaluate the capacity of the STHIRP-1 program, mainly in relation to the thermal model, the new fuel concept was analyzed. The results were compared with those performed with the VIPRE code presented in the reference document.

Optimized High-Temperature Irradiation-Resistant Thermocouple for Fast-Response Measurements

The high-temperature irradiation-resistant thermocouple is the only temperature probe proven to withstand the high-temperature (>1290°C), high-radiation (a fluence of up to ∼1 × 1021 n/cm2) environments of nuclear reactor fuel design testing and/or over-temperature accident conditions. This report describes the improved performance of a molybdenum and niobium thermocouple by utilizing a coaxial design (i.e., a single wire grounded to the outer sheath). This optimized high-temperature irradiation-resistant thermocouple features a simplified design yet allows for more robust individual components. The niobium and molybdenum thermoelements can be used interchangeably in either the sheath or wire, depending on the intended application. Via a plunge test in flowing water, the response time of the coaxial build of the high-temperature irradiation-resistant thermocouple was determined to be 30x faster than that of the comparable ungrounded type-K thermocouples, and 10x faster than the grounded type-K thermocouples and traditional ungrounded high-temperature irradiation-resistant thermocouples (i.e., two-wire configurations). Furthermore, by capitalizing on the coaxial design, a multi-core high-temperature irradiation-resistant probe with multiple “single-pole” wires along the length of the sheath was proven feasible. This multi-core, thermocouple design was dubbed a “demicouple.” The high-temperature irradiation-resistant demicouple is primarily applied during fuel experiments to record multiple fuel-pin centerline temperature measurements using a single compact sensor. Furthermore, the shared “common” leg between demicouple junctions reduces error propagation in secondary measurements such as temperature differentials.

Advanced fuels from ethanol – a superstructure optimization approach

Advanced fuel design through integration of chemistries leading to different components: alcohols (blue); ethers (green); and olefins, parafins, and aromatics (yellow).