The primary metric for the viability of these next generation nuclear power plants will be the cost of generated electricity. One important component in achieving these objectives is the development of power conversion technologies that maximize the electrical power output of these advanced reactors for a given thermal power. More efficient power conversion systems can directly reduce the cost of nuclear generated electricity and therefore advanced power conversion cycle research is an important area of investigation for the Generation IV Program. Brayton cycles using inert or other gas working fluids, have the potential to take advantage of the higher outlet temperature range of Generation IV systems and allow substantial increases in nuclear power conversion efficiency, and potentially reductions in power conversion system capital costs compared to the steam Rankine cycle used in current light water reactors. For the Very High Temperature Reactor (VHTR), Helium Brayton cycles which can operate in the 900 to 950 C range have been the focus of power conversion research. Previous Generation IV studies examined several options for He Brayton cycles that could increase efficiency with acceptable capital cost implications. At these high outlet temperatures, Interstage Heating and Cooling (IHC) was shown to provide significant efficiency improvement (a few to 12%) but required increased system complexity and therefore had potential for increased costs. These scoping studies identified the potential for increased efficiency, but a more detailed analysis of the turbomachinery and heat exchanger sizes and costs was needed to determine whether this approach could be cost effective. The purpose of this study is to examine the turbomachinery and heat exchanger implications of interstage heating and cooling configurations. In general, this analysis illustrates that these engineering considerations introduce new constraints to the design of IHC systems that may require different power conversion configurations to take advantage of the possible efficiency improvement. Very high efficiency gains can be achieved with the IHC approach, but this can require large low pressure turbomachinery or heat exchanger components, whose cost may mitigate the efficiency gain. One stage of interstage cooling is almost always cost effective, but careful optimization of system characteristics is needed for more complex configurations. This report summarizes the primary factors that must be considered in evaluating this approach to more efficient cycles, and the results of the engineering analysis performed to explore these options for Generation IV high temperature reactors.
Sandia National Laboratories is investigating advanced Brayton cycles using supercritical working fluids for use with solar, nuclear or fossil heat sources. The focus of this work has been on the supercritical CO{sub 2} cycle (S-CO2) which has the potential for high efficiency in the temperature range of interest for these heat sources, and is also very compact, with the potential for lower capital costs. The first step in the development of these advanced cycles was the construction of a small scale Brayton cycle loop, funded by the Laboratory Directed Research & Development program, to study the key issue of compression near the critical point of CO{sub 2}. This document outlines the design of the small scale loop, describes the major components, presents models of system performance, including losses, leakage, windage, compressor performance, and flow map predictions, and finally describes the experimental results that have been generated.
An initial version of a Systems Dynamics (SD) modeling framework was developed for the analysis of a broad range of energy technology and policy questions. The specific question selected to demonstrate this process was 'what would be the carbon and import implications of expanding nuclear electric capacity to provide power for plug in hybrid vehicles?' Fifteen SNL SD energy models were reviewed and the US Energy and Greenhouse gas model (USEGM) and the Global Nuclear Futures model (GEFM) were identified as the basis for an initial modeling framework. A basic U.S. Transportation model was created to model U.S. fleet changes. The results of the rapid adoption scenario result in almost 40% of light duty vehicles being PHEV by 2040 which requires about 37 GWy/y of additional electricity demand, equivalent to about 25 new 1.4 GWe nuclear plants. The adoption rate of PHEVs would likely be the controlling factor in achieving the associated reduction in carbon emissions and imports.
This report summarizes the measurements and model predictions for a series of tests supported by the U.S. Department of Energy that were performed using the recently constructed Sandia Brayton Loop (SBL-30). From the test results we have developed steady-state power operating curves, controls methodologies, and transient data for normal and off-normal behavior, such as loss of load events, and for decay heat removal conditions after shutdown. These tests and models show that because the turbomachinery operates off of the temperature difference (between the heat source and the heat sink), that the turbomachinery can continue to operate (off of sensible heat) for long periods of time without auxiliary power. For our test hardware, operations up to one hour have been observed. This effect can provide significant operations and safety benefits for nuclear reactors that are coupled to a Brayton cycles because the operating turbomachinery continues to provide cooling to the reactor. These capabilities mean that the decay-heat removal can be accommodated by properly managing the electrical power produced by the generator/alternator. In some conditions, it may even be possible to produce sufficient power to continue operating auxiliary systems including the waste heat circulatory system. In addition, the Brayton plant impacts the consequences of off-normal and accident events including loss of load and loss of on-site power. We have observed that for a loss of load or a loss of on-site power event, with a reactor scram, the transient consists initially of a turbomachinery speed increase to a new stable operating point. Because the turbomachinery is still spinning, the reactor is still being cooled provided the ultimate heat sink remains available. These highly desirable operational characteristics were observed in the Sandia Brayton loop. This type of behavior is also predicted by our models. Ultimately, these results provide the designers the opportunity to design gas cooled reactor Brayton plants such that the system is inherently capable of dealing with a variety of off-normal and accident conditions.
American Nuclear Society Embedded Topical Meeting - 2007 International Topical Meeting on Safety and Technology of Nuclear Hydrogen Production, Control, and Management
As part of the US DOE Nuclear Hydrogen Initiative, Sandia National Laboratories is designing and constructing a process for the conversion of sulfuric acid to produce sulfur dioxide. This process is part of the thermochemical Sulfur-Iodine (S-I) cycle that produces hydrogen from water. The Sandia process will be integrated with other sections of the S-I cycle in the near future to complete a demonstration-scale S-I process. In the Sandia process, sulfuric acid is concentrated by vacuum distillation and then catalytically decomposed at high temperature (850°C) to produce sulfur dioxide, oxygen and water. Major problems in the process, corrosion, and failure of high-temperature connections of process equipment, have been eliminated through the development of an integrated acid decomposer constructed of silicon carbide. The unit integrates acid boiling, superheating and decomposition into a single unit operation and provides for exceptional heat recuperation. The design of acid decomposition process, the new acid decomposer, other process units, and materials of construction for the process are described and discussed.
The Global Nuclear Energy Partnership fuels development program is currently developing metallic, oxide, and nitride fuel forms as candidate fuels for an Advanced Burner Reactor. The Advance Burner Reactor is being designed to fission actinides efficiently, thereby reducing the long-term storage requirements for spent fuel repositories. Small fuel samples are being fabricated and evaluated with different transuranic loadings and with extensive burnup using the Advanced Test Reactor. During the next several years, numerous fuel samples will be fabricated, evaluated, and tested, with the eventual goal of developing a transmuter fuel database that supports the down selection to the most suitable fuel type. To provide a comparative database of safety margins for the range of potential transmuter fuels, this report describes a plan to conduct a set of early transient tests in the Annular Core Research Reactor at Sandia National Laboratories. The Annular Core Research Reactor is uniquely qualified to perform these types of tests because of its wide range of operating capabilities and large dry central cavity which extents through the center of the core. The goal of the fuels testing program is to demonstrate that the design and fabrication processes are of sufficient quality that the fuel will not fail at its design limit--up to a specified burnup, power density, and operating temperature. Transient testing is required to determine the fuel pin failure thresholds and to demonstrate that adequate fuel failure margins exist during the postulated design basis accidents.
This report describes criticality benchmark experiments containing rhodium that were conducted as part of a Department of Energy Nuclear Energy Research Initiative project. Rhodium is an important fission product absorber. A capability to perform critical experiments with low-enriched uranium fuel was established as part of the project. Ten critical experiments, some containing rhodium and others without, were conducted. The experiments were performed in such a way that the effects of the rhodium could be accurately isolated. The use of the experimental results to test neutronics codes is demonstrated by example for two Monte Carlo codes. These comparisons indicate that the codes predict the behavior of the rhodium in the critical systems within the experimental uncertainties. The results from this project, coupled with the results of follow-on experiments that investigate other fission products, can be used to quantify and reduce the conservatism of spent nuclear fuel safety analyses while still providing the necessary level of safety.
The goal of the Very High Efficiency Reactor study was to develop and analyze concepts for the next generation of nuclear power reactors. The next generation power reactor should be cost effective compared to current power generation plant, passively safe, and proliferation-resistant. High-temperature reactor systems allow higher electrical generating efficiencies and high-temperature process heat applications, such as thermo-chemical hydrogen production. The study focused on three concepts; one using molten salt coolant with a prismatic fuel-element geometry, the other two using high-pressure helium coolant with a prismatic fuel-element geometry and a fuel-pebble element design. Peak operating temperatures, passive-safety, decay heat removal, criticality, burnup, reactivity coefficients, and material issues were analyzed to determine the technical feasibility of each concept.