Numerical Studies of Axial Fuel Shuffling

Materials Today, 2010

2007

The once-through fuel cycle strategy in the United States for the past six decades has resulted in an accumulation of Light Water Reactor (LWR) Spent Nuclear Fuel (SNF). This SNF contains considerable amounts of transuranic (TRU) elements that limit the heat and dose capacity of the current planned repository strategy. A possible way of maximizing the utilization of the repository is to separate the TRU from the LWR SNF through a process such as UREX+1a, and convert it into fuel for a fast-spectrum Advanced Burner Reactor (ABR). The key advantage in this scenario is the assumption that recycling of TRU in the ABR (through pyroprocessing or some other approach), along with a low capture-to-fission probability in the fast reactor's high-energy neutron spectrum, can effectively decrease the decay heat and toxicity of the waste being sent to the repository. The decay heat and toxicity reduction can thus minimize the need for multiple repositories. This report summarizes the work performed by the fuel cycle analysis group at the Idaho National Laboratory (INL) to establish the specific technical capability for performing fast reactor fuel cycle analysis and its application to a high-priority ABR concept. The high-priority ABR conceptual design selected is a metallic-fueled, 1000 MWth SuperPRISM (S-PRISM)-based ABR with a conversion ratio of 0.5. Results from the analysis showed excellent agreement with reference values. The independent model was subsequently used to study the effects of excluding curium from the transuranic (TRU) external feed coming from the LWR SNF and recycling the curium produced by the fast reactor itself through pyroprocessing. Current studies to be published this year focus on analyzing the effects of different separation strategies as well as heterogeneous TRU target systems.

Nuclear Engineering and Technology, 2014

Energies, 2015

In this paper, the mechanism of traveling wave reactors (TWRs) is investigated from the mathematical physics point of view, in which a stationary fission wave is formed by radial fuel drifting. A two dimensional cylindrically symmetric core is considered and the fuel is assumed to drift radially according to a continuous fuel shuffling scheme. A one-group diffusion equation with burn-up dependent macroscopic coefficients is set up. The burn-up dependent macroscopic coefficients were assumed to be known as functions of neutron fluence. By introducing the effective multiplication factor keff, a nonlinear eigenvalue problem is formulated. The 1-D stationary cylindrical coordinate problem can be solved successively by analytical and numerical integrations for associated eigenvalues keff. Two representative 1-D examples are shown for inward and outward fuel drifting motions, respectively. The inward fuel drifting has a higher keff than the outward one. The 2-D eigenvalue problem has to be solved by a more complicated method, namely a pseudo time stepping iteration scheme. Its 2-D asymptotic solutions are obtained together with certain eigenvalues keff for several fuel inward drifting speeds. Distributions of the neutron flux, the neutron fluence, the infinity multiplication factor kinf and the normalized power are presented for two different drifting speeds.

2013

7 Abstract—Questions regarding the feasibility of fusion power are examined, taking into account fuel cycles and breeding reactions, energy balance and reactor conditions, approaches to fusion, magnetic confinement, magneto hydro dynamic instabilities, micro instabilities, and the main technological problems which have to be solved. Basic processes and balances in fusion reactors are considered along with some aspects of the neutronics in fusion reactors, the physics of neutral beam heating, plasma heating by relativistic electrons, radiofrequency heating of fusion plasmas, adiabatic compression and ignition of fusion reactors, dynamics and control of fusion reactors, and aspects of thermal efficiency and waste heat. Attention is also given to fission-fusion hybrid systems, inertial-confinement fusion systems, the radiological aspects of fusion reactors, design considerations of fusion reactors, and a comparative study of the approaches to fusion power. The nuclear fuel cycle, also ...

Annals of Nuclear Energy, 2017

Fast reactor concept like the 2400 MW GCFR is considered a promising candidate for future sustainable and economic nuclear power systems. It is a highly innovative system with advanced geometry and fuel materials (Ceramic fuel pellets of mixed uranium-plutonium carbide within fuel pins). The active fuel region of this reactor is divided radial to two core zones namely, inner core fuel assemblies (IC), and outer core fuel assemblies (OC). In the present study Monte Carlo transport code, MCNPX is used to design heterogeneous and homogeneous models for inner and outer core fuel assembly. The deigned models have been used to simulate the neutronic behavior and fuel transmutation during open cycle operation and to study the effect of Homogeneity inside the fuel assembly. The results of heterogeneous and homogeneous models were compared with each other and good agreement was found.

Nuclear Engineering and Technology, 2018

The aim of this paper is to determine neutronic performances of the light water reactor (LWR) spent fuel mixed with fertile thorium fuel in a FFHR. Time dependent three dimensional calculations for major technical data, such as blanket energy multiplication, tritium breeding ratio, cumulative fissile fuel enrichment and burnup have been performed by using Monte Carlo Neutron-Particle Transport code MCNP5 1.4, coupled with a novel interface code MCNPAS, which is developed by our research group. A self-sustaining tritium breeding ratio (TBR>1.05) has been kept throughout the calculations. The study has shown that the fissile fuel quality will be improved in the course of the transmutation of the LWR spent in the FFHR. The latter has gained the reusable fuel enrichment level conventional LWRs between one and two years. Furthermore, LWR spent fuel-thorium mixture provides higher burn-up values than in light water reactors.

2007

The urgency for addressing repository impacts has grown in the past few years as a result of Spent Nuclear Fuel (SNF) accumulation from commercial nuclear power plants. One obvious path that has been explored by many is to eliminate the transuranic (TRU) inventory from the SNF thus reducing the need for additional long term repository storage sites. One strategy for achieving this is to burn the separated TRU elements in the currently operating U.S. Light Water Reactor (LWR) fleet. Many studies have explored the viability of this strategy by loading a percentage of LWR cores with TRU in the form of either Mixed Oxide (MOX) fuels or Inert Matrix Fuels (IMF). A task was undertaken at INL to establish specific technical capabilities to perform neutronics analyses in order to further assess several key issues related to the viability of thermal recycling. The initial computational study reported here is focused on direct thermal recycling of IMF fuels in a heterogeneous Pressurized Water Reactor (PWR) bundle design containing Plutonium, Neptunium, Americium, and Curium (IMF-PuNpAmCm) in a multipass strategy using legacy 5 year cooled LWR SNF. In addition to this initial high-priority analysis, three other alternate analyses with different TRU vectors in IMF pins were performed.

2017

Axial burnup is an important factor in criticality safety and thermo-hydraulic calculation. In the present study, the effect of axial distribution of burnup on power distribution and isotope inventory were evaluated using combination of the WIMSD-5B and MCNP5 codes for PWR fuel rods assembly. The fuel rods of assembly were divided into 5, 8 and 10 zones. The system was normalized to the total thermal power 16.89172 MWth, which was derived using the fuel assembly load of 458.599 kg. The MCNP5 code has been used to perform three dimensional neutron physics analysis while WIMSD-5B was used for generation of number densities at various stages of fuel burnup. The results were compared with data from the Takahama-3 Benchmark. The calculated-to-experimental and previous calculated results for ratios of U-235, U-236, U-238 and many important nuclides show good agreement. Both results for the isotopic inventory and the power distribution emphasize the importance of considering the axial vari...

Progress in Nuclear Energy, 2011

An axial fuel shuffling strategy is proposed based on the mechanism of the nuclear fission traveling wave and implemented numerically in the calculation for a supercritical water cooled fast reactor (SCWFR). The ERANOS code is adopted to perform the neutronics and burn-up calculations, and the calculation scheme for axial fuel shuffling and coolant density coupling are set up. The parametric studies of a typical PWR with TheU and UePu ( 235 U instead of 239 Pu) conversions by burn-up and k eff calculations indicate that the breeding effects only exist in configurations with very low water content and the conversion or breeding becomes worse as the initial enrichment is increasing. The shuffling calculations for the 1-D SCWFR model described in this paper brought about some interesting results for a certain range of water content. The results indicate that the non-enriched fresh fuel is not possible for both TheU and U ePu conversions. As could be expected due to the h-values of the main fissile isotopes 233 U and ( 235 U, 239 Pu), respectively, the TheU conversion needs a lower enrichment, and results in a slightly higher burn-up than the UePu conversion. The asymptotic power density distribution of the TheU conversion is broader and lower than that of the UePu conversion. By reducing the water volume fraction, an increased burn-up can be achieved with correspondingly reduced fuel shuffling speed and reduced initial enrichment. Furthermore, the steady state calculations for the asymptotic state show that the TheU conversion is superior to the UePu one concerning SCWFR safety aspects, where the absolute value of the Doppler constant is larger and the coolant feedback is negative for the TheU conversion, while the coolant feedback is positive for the UePu one.