Material accountancy in an electrometallurgical Fuel Conditioning Facility

The Fuel Conditioning Facility (FCF) treats spent nuclear fuel using an electrometallurgical process that separates the uranium from the fission products, sodium thermal bond and cladding materials. Material accountancy is necessary at FCF for two reasons: first, it provides a mechanism for detecting a potential loss of nuclear material for safeguards and security; second, it provides a periodic check of inventories to ensure that processes and material are under control. By weighing material entering and leaving a process, and using sampling results to determine composition, an inventory difference (ID) results when the measured inventory is compared to the predicted inventory. The ID and its uncertainty, based on error propagation, determines the degree of assurance that an operation proceeded according to expectations. FCF uses the ID calculation in two ways: closeout, which is the ID and uncertainty for a particular operational step, and material accountancy, which determines an ID and its associated uncertainty for a material balance area through several operational steps. Material accountancy over the whole facility for a specified time period assists in detecting diversion of nuclear material. Data from depleted uranium operations are presented to illustrate the method used in FCF

Current Comparison of Advanced Fuel Cycle Options

The nuclear fuel cycle includes mining, enrichment, nuclear power plants, recycling (if done), and residual waste disposition. The U.S. Advanced Fuel Cycle Initiative (AFCI) has four program objectives to guide research on how best to glue these pieces together, as follows: waste management, proliferation resistance, energy recovery, and systematic management/economics/safety. We have developed a comprehensive set of metrics to evaluate fuel cycle options against the four program objectives. The current list of metrics is long-term heat, long-term dose, radiotoxicity and weapons usable material. This paper describes the current metrics and initial results from comparisons made using these metrics. The data presented were developed using a combination of “static” calculations and a system dynamic model, DYMOND. In many cases, we examine the same issue both dynamically and statically to determine the robustness of the observations. All analyses are for the U.S. reactor fleet. This work aims to clarify many of the issues being discussed within the AFCI program, including Inert Matrix Fuel (IMF) versus Mixed Oxide (MOX) fuel, single-pass versus multi-pass recycling, thermal versus fast reactors, and the value of separating cesium and strontium. The results from a series of dynamic simulations evaluating these options are included in this report. The model interface includes a few “control knobs” for flying or piloting the fuel cycle system into the future. The results from the simulations show that the future is dark (uncertain) and that the system is sluggish with slow time response times to changes (i.e., what types of reactors are built, what types of fuels are used, and the capacity of separation and fabrication plants). Piloting responsibilities are distributed among utilities, government, and regulators, compounding the challenge of making the entire system work and respond to changing circumstances. We identify four approaches that would increase our chances of a sustainable fuel cycle system: (1) have a recycle strategy that could be implemented before the 2030-2050 approximate period when current reactors retire so that replacement reactors fit into the strategy, (2) establish an option such as multi-pass blended-core IMF as a downward Pu control knob and accumulate waste management benefits early, (3) establish fast reactors with flexible conversion ratio as a future control knob that slowly becomes available if/when fast reactors are added to the fleet, and (4) expand exploration of heterogeneous assemblies and cores, which appear to have advantages such as increased agility. Initial results suggest multi-pass full-core MOX appears to be a less effective way than multi-pass blended core IMF to manage the fuel cycle system because it requires higher TRU throughput while accruing waste management benefits at a slower rate. Single-pass recycle approaches for LWRs do not meet AFCI program objectives and could be considered a “dead end.” We did not study the Very High Temperature Reactor (VHTR). Fast reactors appear to be effective options but a significant number of fast reactors must be deployed before the benefit of such strategies can be observed

Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits

Assessment of the SE2-ANL code using EBR-II temperature measurements

The SE2-ANL code is a modified version of the SUPERENERGY-2 code [1]. This code is used at Argonne National Laboratory (ANL) to compute the core-wide temperature profiles in Liquid Metal Reactor (LMR) cores. The accuracy of this code has recently been tested by comparing the predicted temperatures with measured values in the Experimental Breeder Reactor R (EBR-II). The detailed temperature distributions in two experimental subassemblies and the mixed mean subassembly outlet temperatures were used in this validation study. The SE2-ANL predictions were found to agree well with measured values. It was also found that SE2-ANL yields results with accuracy comparable to the more detailed COBRA-WC [2] calculations at much lower computational cost

Development of fibre channel disk clusters. Final report for period September 2, 1998 - March 17, 1999

This report documents the accomplishments of a Phase I project whose purpose was to demonstrate feasibility of developing inexpensive and fast data storage using multi-host Fibre Channel disk clusters. In Phase I, a working file system called ZFS was developed and tested. The ZFS approach was designed to be suited for high energy physics applications, but is general and flexible enough to be useful for other high-volume applications. The ZFS approach, which borrows from the networking concept of cut-through routing, uses Linux boxes and disk clusters in a Fibre Channel--Arbitrated Loop architecture. In ZFS, file locking and other meta-data level operations are carried out over the primary data network, after which all data are sent directly over a Fibre Channel between the workstation and the disk cluster. No intermediate server is required. Substantially higher throughputs than in traditional networked disk architectures have been demonstrated. The ZFS architecture is described and tests of the first implementation of ZFS at Fermilab are discussed. The current system is implemented for Linux and is being optimized for Fermilab's needs, but extensions to other operating systems and other data-intensive applications are clearly foreseen

Temperature effects on interdiffusion of Al and U-Mo under irradiation

A high-energy Xe ion irradiation experiment was conducted to investigate the temperature dependence of interdiffusion in bilayer Al-UMo samples under irradiation. The amount of interdiffusion achieved at a fixed dose with the increase of temperature showed a clear transition at 175 degrees C (with an estimated error in the range of +/- 10 degrees C) from temperature-independent to temperature-dependent behavior. The activation energy derived from the curve of interdiffusion quantity vs. irradiation temperature is 0.77 +/- 0.16 eV. This information has been utilized to understand the temperature effect on the interdiffusion process that occurred at the interfaces of U-Mo particles and the Al matrix in U-Mo/Al dispersion fuels, whose magnitude significantly impacts the fuel's performance. Although this temperature effect was deemed important, it cannot be examined directly using in-pile irradiation data, as fuel temperatures cannot be measured in reactor irradiation and are highly correlated with fission rate and thermal conductivity evolution. To connect the knowledge accumulated from ion irradiation with in-pile irradiation data, simulation of a full-sized U-Mo/Al dispersion fuel plate irradiated in the FUTURE test in the BR2 reactor was performed with the Dispersion Analysis Research Tool (DART), a dispersion fuel performance code. DART is equipped with an interaction or interdiffusion layer (IL) growth correlation formulated to describe the temperature dependence of ion mixing results. The agreement between calculated and measured fuel meat constituent volume fractions and swelling data demonstrated that the temperature effect on in-pile Al-UMo interdiffusion is well captured with the correlation. In this case, the fitted activation energy is 0.70 eV. Considering the uncertainties associated with the ion irradiation data, the activation energy obtained from in-pile data fitting is in accord with that from ion irradiation results. (C) 2020 Elsevier B.V. All rights reserved