The smallest thermal nuclear reactor

There is a growing need for very small nuclear reactors for space applications and as portable high-intensity neutron sources. This technical note investigates the question of what is the smallest possible thermal reactor. It was found that the smallest reactor is a spherically shaped solution of 242mAm(NO3)3 in water. The weight of such a reactor is 4.95 kg with 0.7 kg of 242mAm nuclear fuel. The radius of the reactor in this case is 9.6 cm.

Ultra-thin 242mAm fuel elements in nuclear reactors

There is a growing interest in using 242mAm as a nuclear fuel. The advantages of 242mAm as a nuclear fuel derive from the fact that 242mAm has the highest thermal fission cross section. The thermal capture cross section is relatively low and the number of neutrons per thermal fission is high. These nuclear properties make it possible to obtain nuclear criticality with ultra-thin fuel elements. The possibility of having ultra-thin fuel elements enables the use of these fission products directly, without the necessity of converting their energy to heat, as is done in conventional reactors. There are three options of using such highly energetic and highly ionized fission products. 1. Using the fission products themselves for ionic propulsion. 2. Using the fission products in an MHD generator, in order to obtain electricity directly. 3. Using the fission products to heat a gas up to a high temperature for propulsion purposes. In this work, we are not dealing with a specific reactor design, but only calculating the minimal fuel elements' thickness and the energy of the fission products emerging from these fuel elements. It was found that it is possible to design a nuclear reactor with a fuel element of less than 1 μm of 242mAm. In such a fuel element, 90% of the fission products' energy can escape.

Use of axially graded burnable boron for hot-spot temperature reduction in a pressurized water reactor core

Shortly after the loading of a pressurized water reactor (PWR) core, the axial power distribution in fresh fuel has already attained the characteristic, almost flat shape, typical of burned fuel. At beginning of cycle (BOC), however, the axial distribution is centrally peaked. In assemblies hosting uniform burnable boron rods, this BOC peaking is even more pronounced. A reduction in the axial peaking is today often achieved by shortening the burnable boron rods by some 30 cm at each edge. It is shown that a two-zone grading of the boron rod leads, in a representative PWR cycle, to a reduction of the hot-spot temperature of approximately 70 °C, compared with the nongraded case. However, with a proper three-zone grading of the boron rod, an additional 20 °C may be cut off the hot-spot temperature. Further, with a slightly skewed application of this three-zone grading, an additional 50 °C may be cut off. The representative PWR cycle studied was cycle 11 of the Indian Point 2 station, with a simplification in the number of fuel types and in the burnup distribution. The analysis was based on a complete three-dimensional burnup calculation. The code system was ELCOS, with BOXER as an assembly code for the generation of burnup-dependent cross sections and SILWER as a three-dimensional core code with thermal-hydraulic feedback.

Sensitivity analysis and optimization of the nuclear fuel cycle

A sensitivity study has been conducted to assess the robustness of the conclusions presented in the MIT Fuel Cycle Study. The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycles. The options include limited recycling in LWRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. The analysis allowed optimization of the fast reactor conversion ratio with respect to desired fuel cycle performance characteristics. The following parameters were found to significantly affect the performance of recycling technologies and their penetration over time: Capacity Factors of the fuel cycle facilities, Spent Fuel Cooling Time, Thermal Reprocessing Introduction Date, and incore and Out-of-core TRU Inventory Requirements for recycling technology. An optimization scheme of the nuclear fuel cycle is proposed. Optimization criteria and metrics of interest for different stakeholders in the fuel cycle (economics, waste management, environmental impact, etc.) are utilized for two different optimization techniques (linear and stochastic). Preliminary results covering single and multi-variable and single and multi-objective optimization demonstrate the viability of the optimization scheme.

Numerically stable Monte Carlo-burnup-thermal hydraulic coupling schemes

This paper presents stochastic implicit coupling method intended for use in Monte-Carlo (MC) based reactor analysis systems that include burnup and thermal hydraulic (TH) feedbacks. Both feedbacks are essential for accurate modeling of advanced reactor designs and analyses of associated fuel cycles. In particular, we investigate the effect of different burnup-TH coupling schemes on the numerical stability and accuracy of coupled MC calculations. First, we present the beginning of time step method which is the most commonly used. The accuracy of this method depends on the time step length and it is only conditionally stable. This work demonstrates that even for relatively short time steps, this method can be numerically unstable. Namely, the spatial distribution of neutronic and thermal hydraulic parameters, such as nuclide densities and temperatures, exhibit oscillatory behavior. To address the numerical stability issue, new implicit stochastic methods are proposed. The methods solve the depletion and TH problems simultaneously and use under-relaxation to speed up convergence. These methods are numerically stable and accurate even for relatively large time steps and require less computation time than the existing methods. © 2013 Elsevier Ltd. All rights reserved.