Hussein S. Khalil

Long-Lived Fission Product Transmutation Studies

Abstract A systematic study on long-lived fission products (LLFPs) transmutation has been performed with the aim of devising an optimal strategy for their transmutation in critical or subcritical reactor systems and evaluating impacts on the geologic repository. First, 99Tc and 129I were confirmed to have highest transmutation priorities in terms of transmutability and long-term radiological risk reduction. Then, the transmutation potentials of thermal and fast systems for 99Tc and 129I were evaluated by considering a typical pressurized water reactor (PWR) core and a sodium-cooled accelerator transmutation of waste system. To determine the best transmutation capabilities, various target design and loading optimization studies were performed. It was found that both 99Tc and 129I can be stabilized (i.e., zero net production) in the same PWR core under current design constraints by mixing 99Tc with fuel and by loading CaI2 target pins mixed with ZrH2 in guide tubes, but the PWR option appears to have a limited applicability as a burner of legacy LLFP. In fast systems, loading of moderated LLFP target assemblies in the core periphery (reflector region) was found to be preferable from the viewpoint of neutron economy and safety. By a simultaneous loading of 99Tc and 129I target assemblies in the reflector region, the self-generated 99Tc and 129I as well as the amount produced by several PWR cores could be consumed at a cost of ˜10% increased fuel inventory. Discharge burnups of ˜29 and ˜37% are achieved for 99Tc and 129I target assemblies with an ˜5-yr irradiation period. Based on these results, the impacts of 99Tc and 129I transmutation on the Yucca mountain repository were assessed in terms of the dose rate. The current Yucca Mountain release evaluations do not indicate a compelling need to transmute 99Tc and 129I because the resulting dose rates fall well below current regulatory limits. However, elimination of the LLFP inventory could allow significant relaxation of the waste form and container performance criteria, with associated economic benefits. Therefore, some development of either specialized waste form or transmutation target for the LLFP is prudent, especially considering the potential accumulation of large LLFP inventory with sustained use of nuclear energy into the future.

Blanket Design Studies of a Lead-Bismuth Eutectic-Cooled Accelerator Transmutation of Waste System

Abstract The results of blanket design studies for a lead-bismuth eutectic (LBE)-cooled accelerator transmutation of waste system are presented. These studies focused primarily on achieving two important and somewhat contradictory performance objectives: First, maximizing discharge burnup, so as to minimize the number of successive recycle stages and associated recycle losses, and second, minimizing burnup reactivity loss over an operating cycle, to minimize reduction of source multiplication with burnup. The blanket is assumed to be fueled with a nonuranium metallic dispersion fuel; pyrochemical techniques are used for recycle of residual transuranic (TRU) actinides in this fuel after irradiation. The key system objective of high-discharge burnup is shown to be achievable in a configuration with comparatively high power density and relatively low burnup reactivity loss. System design and operating characteristics that satisfy these goals while meeting key thermal-hydraulic and materials-related design constraints have been preliminarily developed. Results of the performance evaluations indicate that an average discharge burnup of ~27% is achieved with a ~3.5-yr fuel residence time. Reactivity loss over the half-year cycle is 5.3%Δk. The peak fast fluence value at discharge, the TRU fraction in the charged fuel, and the peak coolant velocity are well within the assumed design limits. Owing to its use of nonuranium fuel, this proposed LBE-cooled system can consume light water reactor-discharge TRUs at the maximum rate achievable per unit of fission energy produced (~1.0 g/MWd).

Reconstruction of Pin Power and Burnup Characteristics from Nodal Calculations in Hexagonal-Z Geometry

A reconstruction method is developed for recovering pin burnup characteristics from fuel cycle calculations performed in hexagonal-z geometry using the nodal diffusion option of the DIF3D/REBUS-3 c...

Solution of the mathematical adjoint equations for an interface current nodal formulation

A numerical method for directly computing the mathematical adjoint flux moments and partial currents for the hexagonal-Z geometry interface current nodal formulation in the DIF3D code is described. The new scheme is developed as an alternative to an existing scheme that employs a similarity transformation of the physical adjoint solution to compute the mathematical adjoint. Whereas the existing scheme is rigorous only when the flat transverse-leakage approximation is employed, this new scheme is exact for all leakage approximations in the DIF3D nodal method. in the new scheme, adjoint nodal equations whose form is very similar to that of the forward nodal equations are derived by employing linear combinations of the adjoint partial currents as computational unknowns in the adjoint equations. This enables the use of the forward solution algorithm with only minor modifications for solving the mathematical adjoint equations. By using the new scheme as a reference method, it is shown numerically that while the results computed with the existing scheme are approximate, they are sufficiently accurate for calculations of global and local reactivity changes resulting from coolant voiding in a liquid metal reactor.

Developments in Nuclear Energy Technologies and Nuclear Data Needs

Nuclear data needs can play an important role for innovative nuclear systems. However, in order to establish priority items, a systematic sensitivity/uncertainty analysis must be performed. Same selected examples will be discussed in this paper.

Numerical Benchmarks for Very High-Temperature Reactors Based on the CNPS Critical Experiments

An evaluation of the Compact Nuclear Power Source (CNPS) experiments conducted at Los Alamos National Laboratory in the 1980s has been done using information available in the open literature. The MCNP4C Monte Carlo results for critical test configurations are in good agreement with the experimental values; the keff values are generally within 0.5% of the experimental values. The calculated total and differential rod worths and material worths were also found generally close to experimental values. These good results motivated the utilization of the experimental test data for the specification of two- and three-dimensional numerical benchmark cases that could be used for the verification and validation of core physics codes developed for Very High Temperature Reactor (VHTR) analysis, particularly the deterministic lattice and whole-core physics codes. To define the benchmark cases, the irregular arrangement of channels in the actual CNPS core was simplified to a regular Cartesian geometry arrangement in the benchmark cases, while preserving the important neutronics characteristics of the CNPS. The results of deterministic calculations using the HELIOS/DIF3D code package were compared to MCNP4C results to show the usefulness of the numerical benchmark cases.