As a potential replacement for the National Bureau Standards Reactor (NBSR) at the U.S. National Institute of Standards and Technology (NIST), a conceptual design of a new reactor with a horizontally-split core has recently been studied using low-enriched uranium (LEU) silicide dispersion (USi/Al) fuel. In this paper, the neutronics calculations of the proposed NIST reactor with other two low-enriched U-Mo fuels (U-10Mo monolithic fuel and U-7Mo/Al dispersion fuel) were performed, and the results were compared to that of the USi/Al fuel, with the objective of identifying the best fuel candidate for the reactor cycle length and maximum cold neutron production. To make consistent comparisons, fuel inventories for multi-cycle equilibrium cores were produced for each fuel based on a 30 d reactor cycle at 20 MW thermal power. With its very high uranium density, the potential to load more uranium in the core with U-10Mo monolithic fuel was explored with test cases using an alternate fuel management scheme, a higher power level (30 MW), or a longer cycle (45 d). The research results indicate similar neutronics performance characteristics of the three LEU fuel options in the proposed NIST reactor with the same power level. However, the ability to load more fuel in the reactor with the U-10Mo option allows additional flexibility in the reactor design and could lead to other optimizations that maximize cold neutron production.

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

We review recent work (Briedt et al., 2011., 2012) on a new approach to the formulation and solution of the stochastic inverse parameter determination problem, i.e. determine the random variation of input parameters to a map that matches specified random variation in the output of the map, and then apply the various aspects of this method to the interesting Brusselator model. In this approach, the problem is formulated as an inverse problem for an integral equation using the Law of Total Probability. The solution method employs two steps: (1) we construct a systematic method for approximating set-valued inverse solutions and (2) we construct a computational approach to compute a measure-theoretic approximation of the probability measure on the input space imparted by the approximate set-valued inverse that solves the inverse problem. In addition to convergence analysis, we carry out an a posteriori error analysis on the computed probability distribution that takes into account all sources of stochastic and deterministic error.