Fission Products Distribution in TRISO Coated Fuel Particles Irradiated to 3.22 X 1021 n/cm2 Fast Fluence at 1092°C

Mechanisms by which fission products (especially silver [Ag]) migrate across the coating layers of tristructural isotropic (TRISO) coated fuel particles designed for next generation nuclear reactors have been the subject of a variety of research activities due to the complex nature of the migration mechanisms. This paper presents results obtained from the electron microscopic examination of selected irradiated TRISO coated particles from fuel compact 1-3-1 irradiated in the first Advanced Gas Reactor experiment (AGR-1) that was performed as part of the Next Generation Nuclear Plant (NGNP) project. It is of specific interest to study particles of this compact as they were fabricated using a different carrier gas composition ratio for the SiC layer deposition compared with the baseline coated fuel particles reported on previously. Basic scanning electron microscopy (SEM) and SEM montage investigations of the particles indicate a correlation between the distribution of fission product precipitates and the proximity of the inner pyrolytic carbon (IPyC)-silicon carbide (SiC) interface to the fuel kernel. Transmission electron microscopy (TEM) samples were sectioned by focused ion beam (FIB) technique from the IPyC layer, the SiC layer and the IPyC-SiC interlayer of the coated fuel particle. Detailed TEM and scanning transmission electron microscopy (STEM) coupled with energy dispersive X-ray spectroscopy (EDS) were performed to identify fission products and characterize their distribution across the IPyC and SiC layers in the areas examined. Results indicate the presence of palladium-silicon-uranium (Pd-Si-U), Pd-Si, Pd-U, Pd, U, U-Si precipitates in the SiC layer and the presence of Pd-Si-U, Pd-Si, U-Si, U precipitates in the IPyC layer. No Ag-containing precipitates are evident in the IPyC or SiC layers. With increased distance from the IPyC-SiC interface, there are less U-containing precipitates, however, such precipitates are present across nearly the entire SiC layer.

Effects of the Foil Centering on the Irradiation Performance of U10Mo Alloy Based Monolithic Mini-Plates

Monolithic plate-type fuel is a fuel form being developed for high performance research and test reactors to minimize the use of enriched material. These plate-type fuels consist of a high uranium density LEU foil contained within diffusion barriers and encapsulated within a cladding material. To benchmark this new design, effects of various geometrical and operational variables on irradiation performance have been evaluated. For this work, the effects of fuel foil centering on the thermo-mechanical performance of the mini-plates were studied. To evaluate these effects, a selected plate from RERTR-12 experiments, the Plate L1P756, was considered. The fuel foil was moved within the fuel plate to study the effects of the fuel centering on stress, strain and overall shape of the fuel elements. The thickness of the fuel foil, thickness of the Zr-liners and total thickness of the plate were held constant, except the centerline alignment of the fuel foil. For this, the position of the fuel foil was varied from the center position to a maximum offset corresponding to the minimum allowable aluminum cladding thickness of 0.1524 mm. Results for various offset cases were then compared to each other and to the ideal case of a centered fuel foil. Fabrication simulations indicated that the thermal expansion mismatch results in warping of the fuel plate during fabrication as the fuel plate is cooled from the HIP temperature when the fuel is not centered. Even if the model is constrained during cooling to simulate the rigid HIP can surrounding the fuel plate during cooling, warping is observed when the constraint is removed. Similarly, irradiation simulations revealed that the fuel offset causes virtually all irradiation-induced swelling to occur on the thin-cladding side of the plate. This is observed even for the smallest offset that was considered. The total magnitude of the swelling is approximately same for all offsets values.

Sensitivity Evaluation of the Daily Thermal Predictions of the AGR-2 Experiment in the Advanced Test Reactor

A temperature sensitivity evaluation has been performed for an individual test capsule in the AGR-2 TRISO particle fuel experiment. The AGR-2 experiment is the second in a series of fueled test experiments for TRISO coated fuel particles run in the Advanced Test Reactor at the Idaho National Laboratory. A series of cases were compared to a base case by varying different input parameters in an ABAQUS finite element thermal model. Most input parameters were varied by ±10%, with one parameter ±20%, to show the temperature sensitivity to each parameter. The most sensitive parameters were the outer control gap distance, heat rate in the fuel compacts, and neon gas fraction. The thermal conductivity of the fuel compacts and thermal conductivity of the graphite holder were of moderate sensitivity. The least sensitive parameters were the emissivities of the stainless steel and graphite, along with gamma heat rate in the non-fueled components. Sensitivity calculations were also performed for the fast neutron fluence, which showed a general, but minimal, temperature rise with increasing fluence.

A Supercritical CO2 Brayton Cycle Power Converter for a Sodium-Cooled Fast Reactor Small Modular Reactor

Although a number of power conversion applications have been identified or have even been developed (e.g., waste heat recovery) for supercritical carbon dioxide (S-CO2) cycles including fossil fuel combustors, concentrated solar power (i.e., solar power towers), and marine propulsion, the benefits of S-CO2 Brayton cycle power conversion are especially prominent for applications to nuclear power reactors. In particular, the S-CO2 Brayton cycle is well matched to the Sodium-Cooled Fast Reactor (SFR) nuclear power reactor system and offers significant benefits for SFRs. The recompression closed Brayton cycle is highly recuperated and wants to operate with an approximate optimal S-CO2 temperature rise in the sodium-to-CO2 heat exchangers of about 150 °C which is well matched to the sodium temperature rise through the core that is also about 150 °C. Use of the S-CO2 Brayton cycle eliminates sodium-water reactions and can reduce the nuclear power plant cost per unit electrical power. A conceptual design of an optimized S-CO2 Brayton cycle power converter and supporting systems has been developed for the Advanced Fast Reactor – 100 (AFR-100) 100 MWe-class (250 MWt) SFR Small Modular Reactor (SMR). The AFR-100 is under ongoing development at Argonne National Laboratory (ANL) to target emerging markets where a clean, secure, and stable source of electricity is required but a large-scale power plant cannot be accommodated. The S-CO2 Brayton cycle components and cycle conditions were optimized to minimize the power plant cost per unit electrical power (i.e., $/kWe). For a core outlet temperature of 550 °C and turbine inlet temperature of 517 °C, a cycle efficiency of 42.3 % is calculated that exceeds that obtained with a traditional superheated steam cycle by one percentage point or more. A normal shutdown heat removal system incorporating a pressurized pumped S-CO2 loop slightly above the critical pressure on each of the two intermediate sodium loops has been developed to remove heat from the reactor when the power converter is shut down. Three-dimensional layouts of S-CO2 Brayton cycle power converter and shutdown heat removal components and piping have been determined and three-dimensional CAD drawings prepared. The S-CO2 Brayton cycle power converter is found to have a small footprint reducing the space requirements for components and systems inside of both the turbine generator building and reactor building. The results continue to validate earlier notions about the benefits of S-CO2 Brayton cycle power conversion for SFRs including higher efficiency, improved economics, elimination of sodium-water reactions, load following, and smaller footprint.

High Temperature Intrinsically Safe Nuclear Reactor

Small Modular Reactor (SMR) technologies have been recently included by the DOE as clean energy, a low carbondioxide emitting “alternative energy” source. The objective of this paper is to further the discussion of intrinsically safe nuclear reactors in the context of passive safety design principles and introduction of a novel conceptual reactor design. After a multiple year research study of past fast neutron reactor designs and recent reactor proposals, the following design is the result of analysis of the best concepts discovered. An improved fast reactor of the liquid metal cooled type including a core configuration allowing for only two operational states, “Power” or “Rest”. The flow of the primary cooling fluid suspends the core in the “Power” state, with sufficient flow to remove the heat to an intermediate heat exchanger during normal operation. This design utilizes the force of gravity to shut down the reactor after any loss of coolant flow, either a controlled reactor shut down or a Loss of Coolant Accident (LOCA) event, as the core is controlled via dispersion of fuel elements. Electromagnetic pumps incorporating automatic safety electrical cut-offs are employed to shutdown the primary cooling system to disassemble the core to the “Rest” configuration due to a loss of secondary coolant or loss of ultimate heat sink. This design is a hybrid pool-loop pressurized high-temperature reactor unique in its use of a minimum number of components, utilizing no moving mechanical parts, no rotating seals, and no control rods. This defines an elegantly simple Gen IV intrinsically safe nuclear reactor. [Advanced Small Modular Reactor (aSMR)]

Exergy Analysis for Small Modular Reactor Hybrid Energy System

Nuclear hybrid energy systems (NHES) with the capability to store energy will advance the development of renewable energy technologies by providing reliable, non-carbon emitting, and integrated base-load nuclear energy. Small modular reactors (SMRs) will be significant in establishing hybrid energy systems because of their inherent financial advantages over larger commercial reactors; flexible deployment and faster onsite assembly; and ability to closely match required energy needs for industrial process heat applications. An SMR is a thermal energy plant comprised of many complex systems that interact with each other and their surroundings. To study such a system and set appropriate prices for outputs, it is important to assess thermoeconomics or the effective utility and costs of all resources. At its core, thermoeconomics is based upon the quality of energy, or exergy, flowing into and out of each component within a system. Limited research into the thermoeconomics behind SMRs has been performed, leaving an important gap in understanding. This article presents relevant exergetic cost theory and details methods behind an exergy analysis for an SMR-wind-storage system. To perform this analysis, both the physical and economic environments are identified to provide information on how overall system efficiencies and costs may be analyzed. The physical environment incorporates the actual system components, necessary raw materials, and the surroundings or reference environment. The economic environment refers to the upfront installation and operational costs in addition to market prices. In a purely thermodynamic exergy analysis, the exergetic cost may be determined from the physical environment alone and describes the necessary exergy for production to occur. To improve or optimize a system, system efficiency must be balanced with economics to make NHES more competitive and further their development.

Lessons Learned and Improvements in ANL Plant Dynamics Code Simulation of Experimental S-CO2 Loops

Validation of the ANL Plant Dynamics Code with the experimental data from integral S-CO2 cycle facilities has been continued. Several code modifications as well as modeling approaches and assumptions were introduced to improve both the code’s capabilities in modeling the experimental loops and the agreement of the code prediction with the experimental data. The lessons learned from the code improvement and modeling experience important for the validation of the codes with the experimental data from small-scale integral loops are presented.

Severe Accident Related Vulnerabilities, Potential Design Enhancements and Opportunities for International Cooperation in Risk Reduction in Pressurized Heavy Water Reactors

Operating CANDU PHWRs present significant challenges with respect to their ability to mitigate accidents that are beyond the envelope of design basis drafted over 40 years ago. Today, consideration of severe accidents is a public as well as a regulatory requirement whose implementation begs serious reconsideration in an international coordinated effort. The PHWR enhanced vulnerabilities to accidents such as a sustained loss of AC power, as in Fukushima, arise not only out of the inherent design features but also out of the institutional arrangements that surround their licensing. For one, the reactors will, in absence of a containing pressure vessel as in PWRs, put fission product activity directly into the containment, sport multiple potential containment bypass vulnerabilities and produce copious amounts of flammable gases due to presence of large amounts of Zircaloy in fuel channels and carbon steel in feeders. The relatively thin walled, stepped, welded Calandria vessel into which the disassembled core debris will rest has potential to mechanically fail early, causing explosive and energetic interactions of hot debris with enveloping water. This can catastrophically fail the reactor structures. For severe accidents the containments, well designed for design basis accidents, are either small and weak as in single unit plants or unable to practically take any significant over pressure in negative pressure multi-unit reactor buildings that depend upon a single vacuum building, too small for a multi-unit severe accident. The paper presents analytical arguments in support of these observations, lists conclusions from a series of design reviews and discusses development of ROSHNI, a new generation PHWR dedicated computer code package for simulating an unmitigated station blackout scenario. It does not directly address the institutional issues that handicap a potent reduction of the residual risk posed by continued operation of these reactors without serious design upgrades but discusses the regulatory failures in this regard. It introduces ROSHNI, a newly developed severe accident simulation package that models the reactor core in a greatly enhanced detail necessitated by the variability amongst reactor fuel channels. For a single unit CANDU 6 reactor, the code simulates thermal-mechanical degradation of 4,560 fuel bundles in 380 diverse fuel channels individually (for a total of over 70,000 dissimilar fuel entities) and computes source terms into containment of flammable deuterium gas and fission products. A number of questions are raised about differences between Hydrogen source terms and mitigation measures that are being implemented for light water reactors and Deuterium specific reaction kinetics in generation and mitigation that must be clearly differentiated but ignored so far by PHWR operators. A discussion of effectiveness of certain severe accident specific design upgrade measures that have been implemented at some operating plants is also addressed. For example, potential for a smaller than optimum number (for severe accidents) of PARS units to actually cause Deuterium/Hydrogen explosions as an unintended consequence is discussed. Continued reluctance of CANDU utilities to address a long standing issue of inadequacy of reactor overpressure protection is also detailed.

The Effects of Subcooling on Quenching of a Vertical Brass Cylinder With Heating Power

Film boiling is usually induced while a very hot object contacts with a coolant. Such phenomena will deteriorate the heat transfer and degrade the cooling process. Film boiling is of significant concern for the design of an emergency core cooling system after a hypothetical loss of coolant accident happens in a nuclear power plant. Furthermore, after a nuclear power plant is shut down, the fuel rods will continue to release heat due to the decay of fission products. Moreover, the subcooling of coolant might be changed dramatically during the reflood process. Therefore, it is of significant importance and interest to understand the effect of decay heat and subcooling of coolant on the quenching process of a hot object. This study demonstrates the quenching of a vertical brass cylinder without and with heating power of 105W in deionized water with different subcoolings. The diameter and length of the cylinder is 24 mm and 112 mm, respectively. Six K-Type thermocouples are embedded 2mm below the cylinder surface at different axial locations. The experimental results reveal that, with heating power of 105W, the duration of film boiling becomes much larger than the case without heating power under the same subcooling condition. Besides, the duration of film boiling decreases with increasing subcooling. This study also reveals that the Leidenfrost temperature increases significantly with increasing the subcooling with or without heating power. Significantly, a stable film boiling with approximately constant wall temperature can be sustained in saturated water. This is of significant concern for nuclear safety.