Mary Lou Dunzik-Gougar

Modeling the Nuclear Fuel Cycle

Abstract A review of existing analysis codes for nuclear fuel cycle systems was performed to determine if any existing codes meet technical and functional requirements defined for a U.S. national program supporting the global and domestic assessment, development, and deployment of nuclear energy systems. The program would be implemented using an interconnected architecture of different codes ranging from the fuel cycle analysis code, which is the subject of the review, to fundamental physical and mechanistic codes. Four main functions are defined for the code. Function 1 is the ability to characterize and deploy individual fuel cycle facilities and reactors in a simulation while discretely tracking material movements. Function 2 is the capability to perform an uncertainty analysis for each element of the fuel cycle and an aggregate uncertainty analysis. Function 3 is the inclusion of an optimization engine able to optimize simultaneously across multiple objective functions. Function 4 is open and accessible code software and documentation to aid in collaboration between multiple entities and to facilitate software updates. Existing codes, categorized as annualized or discrete fuel tracking codes, were assessed according to the four functions and associated requirements. These codes were developed by various government, education, and industrial entities to fulfill particular needs. In some cases, decisions were made during code development to limit the level of detail included in a code to ease its use or to focus on certain aspects of a fuel cycle to address specific questions. The review revealed that while no two of the codes are identical, they all perform many of the same basic functions. No code was able to perform defined function 2 or several requirements of functions 1 and 3. Based on this review, it was concluded that the functions and requirements will be met only with development of a new code, referred to as GENIUS.

LIMITED OXIDATION OF IRRADIATED GRAPHITE WASTE TO REMOVE SURFACE CARBON-14

Large quantities of irradiated graphite waste from graphite-moderated nuclear reactors exist and are expected to increase in the case of High Temperature Reactor (HTR) deployment [1,2]. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 ( 14 C), with a half-life of 5730 years. Fachinger et al. [2] have demonstrated that thermal treatment of irradiated graphite removes a significant fraction of the 14 C, which tends to be concentrated on the graphite surface. During thermal treatment, graphite surface carbon atoms interact with naturally adsorbed oxygen complexes to create COx gases, i.e. “gasify” graphite. The effectiveness of this process is highly dependent on the availability of adsorbed oxygen compounds. The quantity and form of adsorbed oxygen complexes in pre- and post-irradiated graphite were studied using Time of Flight Secondary Ion Mass Spectrometry (ToF-SIMS) and Xray Photoelectron Spectroscopy (XPS) in an effort to better understand the gasification process and to apply that understanding to process optimization. Adsorbed oxygen fragments were detected on both irradiated and unirradiated graphite; however, carbon-oxygen bonds were identified only on the irradiated material. This difference is likely due to a large number of carbon active sites associated with the higher lattice disorder resulting from irradiation. Results of XPS analysis also indicated the potential bonding structures of the oxygen fragments removed during surface impingement. Ester- and carboxyl- like structures were predominant among the identified oxygen-containing fragments. The indicated structures are consistent with those characterized by Fanning and Vannice [3] and later incorporated into an oxidation kinetics model by El-Genk and Tournier [4]. Based on the predicted desorption mechanisms of carbon oxides from the identified compounds, it is expected that a majority of the graphite should gasify as carbon monoxide (CO) rather than carbon dioxide (CO₂). Therefore, to optimize the efficiency of thermal treatment the graphite should be heated to temperatures above the surface decomposition temperature increasing the evolution of CO [4].

Microbial Treatment of Irradiated Graphite for Separation of Radioisotope

The Pebble Bed Modular Reactor is being developed in South Africa. Important for PBMR implementation is a viable strategy for waste management. Irradiated graphite from fuel and structural components is too voluminous for practical treatment with traditional higher level waste methods and too radioactive to recycle. To clean the graphite of radionuclides, a two-step process is being pursued: (1) non-carbon radionuclides (activation products, fission products and actinides) are removed on an elemental basis by a chemical or microbial process. (2) 14 C requires separation at an isotopic level, which would be impractical with established methods (gaseous diffusion or centrifuge). PBMR is investigating a method of isotope separation using biofractionation. Preliminary experiments indicate that microorganisms do separate radioactive 14 C from stable 12 C. An aqueous slurry of 14 C-spiked, powdered graphite was “fed” to the microbes for 15–18 hours. The microbes initially contained only background levels of 14 C, i.e. orders of magnitude less than the slurry. In post-experiment analyses, a sample of the microbes was found to contain approximately twice the amount of 14 C present in the bulk slurry material. Experiments are underway to further quantify and verify these results, which indicate distinct microbial processing mechanisms for 14 C and 12 C. The most current results will be presented.Copyright