Simulation of Ion Irradiation of Nuclear Materials and Comparison with Experiment

Radiation defects generated in various nuclear materials such as Mo and CeO2, used as a surrogate material for UO2, formed by sub-MeV Xe and Kr ion implantations were studied via TRIM and MD codes. Calculated results were compared with defect distributions in CeO2 crystals obtained from experiments by implantation of these ions at the doses of 11017 ions/cm2 at several temperatures. A combination of in situ TEM (Transmission Electron Microscopy) and ex situ TEM experiments on Mo were used to study the evolution of defect clusters during implantation of Xe and Kr ions at energies of 150-700 keV, depending on the experimental conditions. The simulation and irradiation were performed on thin film single crystal materials. The formation of defects, dislocations, and solid-state precipitates were studied by simulation and compared to experiment. Void and bubble formation rates are estimated based on a new mesoscale approach that combines experiment with the kinetic models validated by atomistic and Ab-initio simulations. Various sets of quantitative experimental results were obtained to characterize the dose and temperature effects of irradiation. These experimental results include size distributions of dislocation loops, voids and gas bubble structures created by irradiation

Sodium fast reactor fuels and materials : research needs.

An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist

Fuel Cycle Scenario Definition, Evaluation, and Trade-offs

This report 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, potential need for transmutation of technetium and iodine, and the value of separating cesium and strontium. It documents most of the work produced by INL, ANL, and SNL personnel under their Simulation, Evaluation, and Trade Study (SETS) work packages during FY2005 and the first half of FY2006. This report represents the first attempt to calculate a full range of metrics, covering all four AFCI program objectives - waste management, proliferation resistance, energy recovery, and systematic management/economics/safety - 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 is a technical report, not aimed at a policy-level audience. A wide range of options are studied to provide the technical basis for identifying the most attractive options and potential improvements. Option improvement could be vital to accomplish before the AFCI program publishes definitive cost estimates. Information from this report will be extracted and summarized in future policy-level reports. Many dynamic simulations of deploying those options are included. There are few "control knobs" for flying or piloting the fuel cycle system into the future, even though it is dark (uncertain) and controls are sluggish with slow time response: 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 ability to pilot the 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 plutonium 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 blended assemblies and cores, which appear to have advantages and 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 more slowly accruing waste management benefits. Single-pass recycle approaches for LWRs (we did not study the VHTR) do not meet AFCI program objectives and could be considered a "dead end". 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

Sodium fast reactor safety and licensing research plan. Volume II.

Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I

Modeling the Nuclear Fuel Cycle: Agent Based Approach

International audienceSystem dynamics approach was previously used to simulate the dynamics of the nuclear fuel cycle and associated infrastructure deployments. It was found to be a useful paradigm that is appropriate for simulations of this system given the nuclear fuel cycle’s inherent mass flows, process time delays, and feedback loops. Agent based simulation approach is currently considered as an alternative or complimentary approach to system dynamics for this type of simulations. The move towards agent based simulations is motivated by the needs to expand the options for fuel cycle simulation and to explore further synergies between the different components and players that affect the behavior of the fuel cycle developments. This paper explores the applicability of both approaches to the nuclear fuel cycle simulation and discuss an agent based model of the fuel cycle, SINDA model, that can be further developed in the future to explore expanded and more realistic fuel cycle deployment scenarios

VISION: Verifiable Fuel Cycle Simulation Model

The nuclear fuel cycle is a very complex system that includes considerable dynamic complexity as well as detail complexity. In the nuclear power realm, there are experts and considerable research and development in nuclear fuel development, separations technology, reactor physics and waste management. What is lacking is an overall understanding of the entire nuclear fuel cycle and how the deployment of new fuel cycle technologies affects the overall performance of the fuel cycle. The Advanced Fuel Cycle Initiative’s systems analysis group is developing a dynamic simulation model, VISION, to capture the relationships, timing and delays in and among the fuel cycle components to help develop an understanding of how the overall fuel cycle works and can transition as technologies are changed. This paper is an overview of the philosophy and development strategy behind VISION. The paper includes some descriptions of the model and some examples of how to use VISION

Morphological characterization of the fresh ZrN coated UMo powders used in EMPIrE irradiation experiment: A practical approach

The European Mini-Plate Irradiation Experiment (EMPIrE) aims to test the in-pile behavior of various ZrN coated UMo powder batches, produced using different technological processes. UMo(ZrN) particles are typical core-shell systems taking advantages of a dense fissile material, UMo, with a coated ZrN layer acting as a diffusion barrier. The U-7Mo kernels were produced by centrifugal or rotating electrode atomization processes and the ZrN coating was performed by physical vapor deposition or atomic layer deposition processes. A total of ten batches of UMo(ZrN) powders were examined in the as-obtained state, i.e prior to fuel-plate fabrication and before in-pile irradiation. The present investigation gives a characterization of each powder batch mainly in terms of morphological and microstructural features carried out by means of SEM-EDS, LOM, EPMA, AFM and TEM analyses. Digital image processing using ImageJ software was employed to determine several particle size (major axis, minor axis, Feret’s diameters, and equivalent diameter) and shape (aspect ratio, circularity, convexity and concavity) parameters as well as the ZrN deposited layer thickness. The quality of the ZrN layer was examined in terms of surface roughness, grain structure and aggregated habits. Our characterization results draw a detailed portrait of the UMo(ZrN) powders selected for the EMPIrE experiment and allow a classification of the powder batches which is presented as radar chart (Kiviat diagram)

Investigation of High-Energy Ion-Irradiated MA957 Using Synchrotron Radiation under In-Situ Tension

In this study, an MA957 oxide dispersion-strengthened (ODS) alloy was irradiated with high-energy ions in the Argonne Tandem Linac Accelerator System. Fe ions at an energy of 84 MeV bombarded MA957 tensile specimens, creating a damage region ~7.5 μm in depth; the peak damage (~40 dpa) was estimated to be at ~7 μm from the surface. Following the irradiation, in-situ high-energy X-ray diffraction measurements were performed at the Advanced Photon Source in order to study the dynamic deformation behavior of the specimens after ion irradiation damage. In-situ X-ray measurements taken during tensile testing of the ion-irradiated MA957 revealed a difference in loading behavior between the irradiated and un-irradiated regions of the specimen. At equivalent applied stresses, lower lattice strains were found in the radiation-damaged region than those in the un-irradiated region. This might be associated with a higher level of Type II stresses as a result of radiation hardening. The study has demonstrated the feasibility of combining high-energy ion radiation and high-energy synchrotron X-ray diffraction to study materials’ radiation damage in a dynamic manner

Sodium fast reactor fuels and materials : research needs.

An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist