Department Of Energy Office Of Nuclear Energy Research Articles

Development of the Versatile Test Reactor Probabilistic Risk Assessment

The Versatile Test Reactor (VTR) is a fast spectrum test reactor currently being developed in the United States under the direction of the U.S. Department of Energy (DOE), Office of Nuclear Energy (DOE-NE). The mission of the VTR is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The conceptual design of the 300-MW(thermal), sodium-cooled, metallic-fueled, pool-type fast reactor has been led by U.S. national laboratories in collaboration with General Electric-Hitachi and Bechtel National Inc. To facilitate risk-informed design and authorization activities during the conceptual development phase, a conceptual design probabilistic risk assessment (PRA) was performed for the VTR. This paper provides an overview of the development of the VTR conceptual design PRA, including key DOE and industry standards and the PRA analysis approach and structure. In addition, the results of the VTR conceptual design PRA are provided, which include its use within authorization documentation and design decisions, along with important lessons learned during the process. The work reported in the paper is the result of studies supporting a VTR conceptual design, cost, and schedule estimate for DOE-NE to make a decision on procurement. As such, it is preliminary.

Transformational Challenge Reactor preconceptual core design studies

In the nuclear industry, a manufacturing-informed design approach has the potential to yield the most benefit from advanced manufacturing. By leveraging advanced materials, data science, and rapid testing and deployment, manufacturing-informed design can drive down costs and development times, ultimately improving future commercial viability. This approach is being demonstrated in the US Department of Energy Office of Nuclear Energy (DOE-NE) Transformational Challenge Reactor (TCR) program. Preconceptual design activities for TCR have been focused on analyzing and maturing four reactor core design concepts: two fast-spectrum and two thermal-spectrum systems. The designs were iteratively modified and analyzed, and subcomponents were manufactured in parallel over weeks instead of months or years. To meet key program initiatives (e.g., timeline and material use), several constraints—including fissile material availability, component availability, materials compatibility, and additive manufacturing capabilities—were factored into the design effort, yielding small cores less than one cubic meter in volume with near-term viability.The TCR program has made significant progress on development of advanced moderator materials such as yttrium hydride, advancing the feasibility of gas-cooled thermal spectrum systems using less than 250 kg of high-assay low enriched uranium (HALEU) and occupying less than 1 m3. Each of the two resulting thermal designs uses a different fuel form: traditional UO2 ceramic fuel and tristructural isotropic (advanced TRISO) fuel particles embedded inside a SiC matrix. Core neutronics and thermal performance for these systems were assessed and summarized. Evaluation of the performance metrics for these two moderated designs has yielded the downselected TCR design: a TRISO-fueled and yttrium hydride moderated gas-cooled reactor.

Modeling and simulation functional needs for molten salt reactor licensing

The modeling and simulation functional needs to license and design molten salt reactor concepts were assessed and an evaluation methodology was developed using a licensing-focused approach considering the modeling and simulation components that contribute to a molten salt reactor safety analysis report. These functional needs were prioritized by identifying the relationships between functional needs and using a quantitative weighting method to identify the most critical components to build a licensing case. This novel weighting method provided a list of functional needs prioritized relative to key advanced reactor licensing outcomes. In addition, functional need inputs were ranked to reveal those that feed the most modeling and simulation components. The findings from this prioritization effort were used to inform the modeling and simulation goals of the US Department of Energy Office of Nuclear Energy (DOE-NE) Molten Salt Reactor campaign.Modeling and simulation needs were grouped into three main categories: (1)source term definition, material accountancy, and radiological consequences, (2)safety analysis, and (3)steady-state neutronics and thermal hydraulics. Five near-term goals were identified among these focus areas that add value: (1) deploying a systems and safeguards analysis computational tool, (2) defining source terms for molten salt reactor analysis and interfacing with the US Nuclear Regulatory Commission on accident progression analysis, (3) developing point designs for liquid-fueled thermal-spectrum and fast-spectrum molten salt reactors, (4) building safety analysis specifications and reference solutions for Molten Salt Reactor Experiment transients, and (5)interfacing with the thermophysical and thermodynamic salt chemistry databases.

Electron Microscopy Observations of Catalyst-Support Interactions in Polymer Electrolyte Membrane Fuel Cells

The activity and durability of both platinum group metal (PGM)-free and low-PGM catalysts rely heavily on the interaction of the active species with the support structure [1]. These carbon-based supports can be tailored in a number of ways to impact the morphology and performance of the electrocatalyst. Visualizing these interactions by scanning transmission electron microscopy (STEM) in two and/or three dimensions can provide valuable multi-scale insight, which can be leveraged towards maximizing catalyst utilization and durability. In this work, quantitative analytical STEM characterization is performed at high spatial resolution on a number of Pt-based and PGM-free catalyst and support systems to elucidate the nature of these critical interactions. The interplay between carbon graphitization and Pt loading was examined in three dimensions by quantitative electron tomography. The initial Pt dispersion and subsequent agglomeration following accelerated stress tests were quantified for high surface area carbon, Vulcan, and low surface-area carbons, with Pt catalyst loadings ranging from 5 to 40 wt.%. Higher degrees of graphitization in the carbon support led to a poor initial Pt dispersion, which in turn resulted in increased agglomeration during cycling. These results showed that in order to improve the mass activity and durability of the corrosion-resistance graphitized carbons, steps must be taken to improve initial Pt dispersion. To this end, a series of nitrogen-doped graphitized carbon supports were synthesized to study possible interactions between the nitrogen-doped surfaces and Pt nanoparticles. Low-voltage, aberration-corrected STEM coupled with electron energy loss spectroscopy (EELS) was used to probe these interactions in an effort to improve Pt dispersion and stability on corrosion-resistance supports. Such catalyst-support interactions are perhaps best epitomized in the atomic-level interplay found in heat-treated metal-nitrogen-carbon (M-N-C) catalysts in fuel cells with PGM-free cathodes [2]. The resulting atomic-level structures derived from zeolitic imidazolate frameworks (ZIFs) and hybrid cyanamide-polyaniline precursors were visualized at the atomic level by the low-voltage, aberration-corrected STEM. The composition of catalytic oxygen reduction reaction (ORR) active sites and other structures explored by STEM-EELS complement electrochemical measurements to further guide synthesis protocols as efforts continue to produce high-performance and durable PGM-free based fuel cells. References L. Du, Y. Shao, J. Sun, G. Yin, J. Liu, Y. Wang, Nano Energy 29 (2016) 314.G. Wu, K. L. More, C. M. Johnston, P. Zelenay, Science 332 (2011) 443. Acknowledgements Research sponsored by the Fuel Cell Technologies Office, Office of Energy Efficiency and Renewable Energy, U.S. Department of Energy (DOE), and through a user project supported by ORNL’s Center for Nanophase Materials Sciences (CNMS), which is a DOE Office of Science User Facility. A portion of this research was performed using instrumentation provided by the U.S. DOE Office of Nuclear Energy, Fuel Cycle R&D Program, and the Nuclear Science User Facilities. Figure 1

United States Department of Energy Severe Accident Research Following the Fukushima Daiichi Accidents

The U.S. Department of Energy (DOE) has played a major role in the U.S. response to the events at Fukushima Daiichi. During the first several weeks following the accident, U.S. assistance efforts were guided by results from a significant and diverse set of analyses. In the months that followed, a coordinated analysis activity aimed at gaining a more thorough understanding of the accident sequence was completed using laboratory-developed, system-level best-estimate accident analysis codes, while a parallel analysis was conducted by U.S. industry. A comparison of predictions for Unit 1 from these two studies indicated significant differences between MAAP and MELCOR results for key plant parameters, such as in-core hydrogen production. On that basis, a crosswalk was completed to determine the key modeling variations that led to these differences. In parallel with these activities, it became clear that there was a need to perform a technology gap evaluation on accident-tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist given the current state of light water reactor (LWR) severe accident research and augmented by insights from Fukushima. In addition, there is growing international recognition that data from Fukushima could significantly reduce uncertainties related to severe accident progression, particularly for boiling water reactors. On these bases, a group of U.S. experts in LWR safety and plant operations was convened by the DOE Office of Nuclear Energy (DOE-NE) to complete technology gap analysis and Fukushima forensics data needs identification activities. The results from these activities were used as the basis for refining DOE-NE's severe accident research and development (R&D) plan. This paper provides a high-level review of DOE-sponsored R&D efforts in these areas, including planned activities on accident-tolerant components and accident analysis methods.

Xenon Recovery from Molybdenum-99 Production

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) sponsors research and development on the recycle of used commercial nuclear fuel as an option for future nuclear fuel cycles that offers increased use of uranium and thorium resources and a possible reduction in the overall cost of nuclear waste management. The two alternatives, direct disposal of used fuel and fuel recycle, are broadly referred to as open and closed fuel cycles. One requirement of a closed fuel cycle is the safe management of radioactive off-gases, which includes 14C, radioiodine and the noble gases, including radioxenon. The longest lived relevant radioxenon isotope is 127Xe; with a half-life of just 36.35 days it is feasible to trap and hold the radioxenon to allow for decay to safe environmental levels. However, the very weak chemical bonds of noble gases, in this case xenon, make them difficult to trap, which led to an extensive DOE-NE study of noble gas adsorption on various molecular sieves as an alternative to costly cryogenics processes. Preliminary results indicate that xenon adsorption at near room temperature on molecular sieves, both synthetic and natural, may have both cost and efficiency advantages over cryogenic processes.Xenon radioisotopes with short half-lives have value in the detection of underground nuclear explosions. However radioxenon released by medical isotope production facilities produces high backgrounds that complicate its use for nuclear explosion detection. Specifically, recovery of 99Mo for medical applications from short-cooled irradiated 235U targets results in the release of some radioxenon. In March 2015, by a unanimous vote of the five permanent members of the United Nations Security Council, commercial producers of 99Mo were asked to decrease and, if possible, eliminate the emissions of radioactive xenon from their production processes. The use of materials currently being evaluated for Xe and Kr capture for use in UNF processing applications may provide a cost effective mechanism to achieve the United Nations Security Council goal.

Nuclear Radiation Tolerance of Single Crystal Aluminum Nitride Ultrasonic Transducer

Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models, (Rempe et al., 2011; Kazys et al., 2005). These efforts are limited by the lack of identified ultrasonic transducer materials capable of long term performance under irradiation test conditions. To address this need, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate the performance of promising magnetostrictive and piezoelectric transducers in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 1021 n/cm2. The irradiation is also supported by a multi-National Laboratory collaboration funded by the Nuclear Energy Enabling Technologies Advanced Sensors and Instrumentation (NEET ASI) program. The results from this irradiation, which started in February 2014, offer the potential to enable the development of novel radiation tolerant ultrasonic sensors for use in Material Testing Reactors (MTRs). As such, this test is an instrumented lead test and real-time transducer performance data is collected along with temperature and neutron and gamma flux data. Hence, results from this irradiation offer the potential to bridge the gap between proven out-of-pile ultrasonic techniques and in-pile deployment of ultrasonic sensors by acquiring the data necessary to demonstrate the performance of ultrasonic transducers. To date, very encouraging results have been attained as several transducers have continued to operate under irradiation. The irradiation is ongoing and will continue to approximately mid-2015.

Irradiation Testing of Ultrasonic Transducers

Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of numerous parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of single, small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. Other efforts include an ultrasonic technique to detect morphology changes (such as crack initiation and growth) and acoustic techniques to evaluate fission gas composition and pressure. These efforts are limited by the lack of existing knowledge of ultrasonic transducer material survivability under irradiation conditions. To address this need, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate promising magnetostrictive and piezoelectric transducer performance in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 10 <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">21</sup> n/cm <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> (E>0.1 MeV). This test will be an instrumented lead test; and real-time transducer performance data will be collected along with temperature and neutron and gamma flux data. By characterizing magnetostrictive and piezoelectric transducer survivability during irradiation, test results will enable the development of novel radiation tolerant ultrasonic sensors for use in Material and Test Reactors (MTRs). The current work bridges the gap between proven out-of-pile ultrasonic techniques and in-pile deployment of ultrasonic sensors by acquiring the data necessary to demonstrate the performance of ultrasonic transducers.

Instrumentation Performance During the TMI-2 Accident

The accident at the Three Mile Island Unit 2 (TMI-2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focused upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this paper. As noted within this paper, several techniques were invoked in the TMI-2 post-accident program to evaluate sensor survivability status and data qualification, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this paper provides recommendations related to sensor survivability and the data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts.

Potentiometric Sensor for Multivalent Ions in High Temperature Molten Salt

Natalie Gese*, Brenda Serrano, Jan-Fong Jue, Guy FredricksonIdaho National Laboratory, Idaho Falls, ID, USAA solid-state potentiometric sensor [1] is being developed for ion selective monitoring of multivalent ion concentrations in LiCl-KCl molten salt. Based on the present chemical and mechanical stability studies it has been demonstrated that a solid electrolyte based sensor composed of polycrystalline ß"- alumina is well suited to aggressive molten salt environments where the sensor would be directly placed.A natural consequence of electrorefining spent nuclear fuel is that during electrochemical oxidation of uranium, active metals and fission products also report to the ternary chloride molten salt comprising LiCl-KCl-UCl3. The UCl3 in the electrolyte is necessary for efficient electrotransport of uranium from the anode to the cathode and for obtaining a high purity cathode product; however, uranium favors the metal phase relative to other actinides [2]. As many batches of spent fuel are processed the UCl3 concentration decreases in the electrolyte. Elements in the spent fuel which are more stable in the molten salt as chlorides than uranium, participate in redox displacement reactions with UCl3. One of the main elements of concern is plutonium which forms PuCl3 at the expense of UCl3 and accumulates in the salt phase. Monitoring PuCl3 in the salt phase is necessary for safeguarding, process efficiency and control, and maintaining regulatory limits. Currently, spectroscopic methods are used to determine the composition of the salt between batches of fuel processed but there is no method in place to monitor actinide concentrations in the salt phase during electrorefining. There are significant time delays which are associated with the transfer, handling and analysis required. A real-time in situ ion selective sensor for plutonium and other actinides would provide many advantages as method for determining plutonium concentration so that electrorefining operations can be further expedited, the salt chemistry can be maintained efficiently, and a method for safeguarding can be employed for monitoring the salt phase.Gadolinium was used as a multivalent ion surrogate for plutonium and other actinides in this study. Ion exchange of mobile sodium ions for gadolinium ions was achieved in polycrystalline Na-ß"-alumina ion conducting ceramics to prepare the sensor material [3]. Annealing was also performed to enhance the mechanical integrity of the Gd-ß"-alumina prepared through the ion exchange process. Conductivity as a function of temperature and the activation energy for ionic conduction was determined with Electrochemical Impedance Spectroscopy (EIS). Ion exchange, chemical stability and mechanical integrity were verified with SEM/EDS analysis. X-ray diffraction (XRD) was used to characterize the phase stability and phase changes in the Gd-ß"-alumina as a result of ion exchange and annealing. The potential between the Pt/Gd-ß"-alumina/Pt cell was measured as function of change in gadolinium concentration in LiCl-KCl. The approach and results obtained in development of this sensor will be presented.This work was supported by the Department of Energy's Office of Nuclear Energy (DOE-NE), Fuel Cycle Research and Development (FCR&D), Materials Protection, Accounting and Control for Transmutation (MPACT). The authors would like to thank Dr. Shelly Lee for creating this research incentive.[1] Fray, D., Metall and Materi Trans B2003, 34 (5), 589-594.[2] Tomczuk, Z.; Ackerman, J. P.; Wolson, R. D.; Miller, W. E., J. Electrochem. Soc.1992, 139 (12), 3523-3528.[3] Dunn, B.; Farrington, G. C., Solid State Ionics 1983, 9-10 (Part 1), 223-225.