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.

The effects of operational and design parameters on the tolerance of ADSR fuel pin cladding to beam interruptions

This paper investigates the effects of design parameters, such as cladding and coolant material choices, and operational phenomena, such as creep and fission product decay heat, on the tolerance of Accelerator Driven Subcritical Reactor (ADSR) fuel pin cladding to beam interruptions. This work aims to provide a greater understanding of the integration between accelerator and nuclear reactor technologies in ADSRs. The results show that an upper limit on cladding operating temperature of 550 °C is appropriate, as higher values of temperature tend to accelerate creep, leading to cladding failure much sooner than anticipated. The effect of fission product decay heat is to reduce significantly the maximum stress developed in the cladding during a beam-trip-induced transient. The potential impact of irradiation damage and the effects of the liquid metal coolant environment on the cladding are discussed. © 2013 Elsevier Ltd. All rights reserved.

Implementation of multi-group cross-section methodology in BGCore MC-depletion code

BGCore is a software package for comprehensive computer simulation of nuclear reactor systems and their fuel cycles. The BGCore interfaces Monte Carlo particles transport code MCNP4C with a SARAF module - an independently developed code for calculating in-core fuel composition and spent fuel emissions following discharge. In BGCore system, depletion coupling methodology is based on the multi-group approach that significantly reduces computation time and allows tracking of large number of nuclides during calculations. In this study, burnup calculation capabilities of BGCore system were validated against well established and verified, computer codes for thermal and fast spectrum lattices. Very good agreement in k eigenvalue and nuclide densities prediction was observed for all cases under consideration. In addition, decay heat prediction capabilities of the BGCore system were benchmarked against the most recent edition of ANS Standard methodology for UO2 fuel decay power prediction in LWRs. It was found that the difference between ANS standard data and that predicted by the BGCore does not exceed 5%.

Muitiobjective pressurized water reactor reload core design by nondominated genetic algorithm search

The design of pressurized water reactor reload cores is not only a formidable optimization problem but also, in many instances, a multiobjective problem. A genetic algorithm (GA) designed to perform true multiobjective optimization on such problems is described. Genetic algorithms simulate natural evolution. They differ from most optimization techniques by searching from one group of solutions to another, rather than from one solution to another. New solutions are generated by breeding from existing solutions. By selecting better (in a multiobjective sense) solutions as parents more often, the population can be evolved to reveal the trade-off surface between the competing objectives. An example illustrating the effectiveness of this novel method is presented and analyzed. It is found that in solving a reload design problem the algorithm evaluates a similar number of loading patterns to other state-of-the-art methods, but in the process reveals much more information about the nature of the problem being solved. The actual computational cost incurred depends: on the core simulator used; the GA itself is code independent.