Publications

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The mission of the Energy Conversion (EC) area of the Advanced Reactor Technology (ART) program is to commercialize the sCO₂ Brayton cycle for Advance Reactors and for the Supercritical Transformational Electric Production (STEP) program. The near-term objective of the EC team efforts is to support the development of a commercially scalable Recompression Closed Brayton Cycle (RCBC) to be constructed for the first STEP demonstration system with the lowest risk possible. This document details the status of technology, policy and market considerations, documentation of gaps and needs, and outlines the steps necessary for the successful development and deployment of commercial sCO₂ Brayton Power Systems along the path to nuclear reactor applications.

The purpose of the ART Energy Conversion (EC) Project is to provide solutions to convert the heat from an advanced reactor to useful products that support commercial application of the reactor designs.

Over the past ten years, the Department of Energy (DOE) has helped to develop components and technologies for the Supercritical Carbon Dioxide (sCO₂) power cycle capable of efficient operation at high temperatures and high efficiency. The DOE Offices of Fossil Energy, Nuclear Energy, and Energy Efficiency and Renewable Energy collaborated in the planning and execution of the sCO₂ Power Cycle Summit conducted in Albuquerque, NM in November 2017. The summit brought together participants from government, national laboratories, research, and industry to engage in discussions regarding the future of sCO₂ Power Cycles Technology. This report summarizes the work involved in summit planning and execution, before, during, and after the event, including the coordination between three DOE offices and technical content presented at the event.

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The near-term objective of the EC team is to establish an operating, commercially scalable Recompression Closed Brayton Cycle (RCBC) to be constructed for the NE - STEP demonstration system (demo) with the lowest risk possible. A systems engineering approach is recommended to ensure adequate requirements gathering, documentation, and mode ling that supports technology development relevant to advanced reactors while supporting crosscut interests in potential applications. A holistic systems engineering model was designed for the ART Energy Conversion program by leveraging Concurrent Engineering, Balance Model, Simplified V Model, and Project Management principles. The resulting model supports the identification and validation of lifecycle Brayton systems requirements, and allows designers to detail system-specific components relevant to the current stage in the lifecycle, while maintaining a holistic view of all system elements.

The supercritical carbon dioxide (S - CO2) Brayton Cycle has gained significant attention in the last decade as an advanced power cycle capab le of achieving high efficiency power conversion. Sandia National Laboratories, with support from the U.S. Department of Energy Office of Nuclear Energy (US DOE - NE), has been conducting research and development in order to deliver a technology that is rea dy for commercialization. There are a wide range of materials related challenges that must be overcome for the success of this technology. At Sandia, recent work has focused on the following main areas: (1) Investigating the potential for system cost re duction through the introduction of low cost alloys in low temperature loop sections, (2) Identifying material options for 10MW RCBC systems, (3) Understanding and resolving turbine degradation, (4) Identifying gas foil bearing behavior in CO 2 , and (5) Ide ntifying the influence of gas chemistry on alloy corrosion. Progress in each of these areas is provided in this report.

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Sandia National Laboratories and General Atomics are pleased to respond to the Advanced Research Projects Agency-Energy (ARPA-e)’s request for information on innovative developments that may overcome various current reactor-technology limitations. The RFI is particularly interested in innovations that enable ultra-safe and secure modular nuclear energy systems. Our response addresses the specific features for reactor designs called out in the RFI, including a brief assessment of the current state of the technologies that would enable each feature and the methods by which they could be best incorporated into a reactor design.

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At Sandia National Laboratories (SNL), The Nuclear Energy Systems Laboratory / Brayton Lab has been established to research and develop subsystems and demonstrate the viability of the closed Brayton cycles (CBC), and in particular, the recompression CBC. The ultimate objective of this program is to have a commercial-ready system available for small modular reactors. For this objective, R&D efforts must demonstrate that, among other things, component and the system behavior is understood and control is manageable, and system performance is predictable. Research activities that address these needs include investigating system responses to various anticipated perturbations, and demonstrating that component and system performance is understood. To these ends, this paper presents system response to a perturbation, and turbomachinery performance results during steady state operation. A long duration test, with an extensive period at steady state, was completed in the simple CBC configuration. During this period, a cooling perturbation was initiated. Data from this test are presented and evaluated to explain the sequence of events following the perturbation. It was found that a cascading series of events ensued, starting with the fluid condensing effect of the cooling perturbation. The explanation of events emphasizes the highly interactive and nonlinear nature of CBC's. The comparisons of measured and predicted turbomachinery performance yielded excellent results and give confidence that the predictive methods originally envisioned for this system work well.

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Small-scale supercritical CO2 demonstration loops are successful at identifying the important technical issues that one must face in order to scale up to larger power levels. The Sandia National Laboratories supercritical CO2 Brayton cycle test loops are identifying technical needs to scale the technology to commercial power levels such as 10 MWe. The small size of the Sandia 1 MWth loop has demonstration of the split flow loop efficiency and effectiveness of the Printed Circuit Heat Exchangers (PCHXs) leading to the design of a fully recuperated, split flow, supercritical CO2 Brayton cycle demonstration system. However, there were many problems that were encountered, such as high rotational speeds in the units. Additionally, the turbomachinery in the test loops need to identify issues concerning the bearings, seals, thermal boundaries, and motor controller problems in order to be proved a reliable power source in the 300 kWe range. Although these issues were anticipated in smaller demonstration units, commercially scaled hardware would eliminate these problems caused by high rotational speeds at small scale. The economic viability and development of the future scalable 10 MWe solely depends on the interest of DOE and private industry. The Intellectual Property collected by Sandia proves that the ~10 MWe supercritical CO2 power conversion loop to be very beneficial when coupled to a 20 MWth heat source (either solar, geothermal, fossil, or nuclear). This paper will identify a commercialization plan, as well as, a roadmap from the simple 1 MWth supercritical CO2 development loop to a power producing 10 MWe supercritical CO2 Brayton loop.

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This report describes the supercritical carbon dioxide (S-CO{sub 2}) direct cycle gas fast reactor (SC-GFR) concept. The SC-GFR reactor concept was developed to determine the feasibility of a right size reactor (RSR) type concept using S-CO{sub 2} as the working fluid in a direct cycle fast reactor. Scoping analyses were performed for a 200 to 400 MWth reactor and an S-CO{sub 2} Brayton cycle. Although a significant amount of work is still required, this type of reactor concept maintains some potentially significant advantages over ideal gas-cooled systems and liquid metal-cooled systems. The analyses presented in this report show that a relatively small long-life reactor core could be developed that maintains decay heat removal by natural circulation. The concept is based largely on the Advanced Gas Reactor (AGR) commercial power plants operated in the United Kingdom and other GFR concepts.

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This Sandia supported research project evaluated the potential improvement that 'condensing' supercritical carbon dioxide (S-CO{sub 2}) power cycles can have on the efficiency of Light Water Reactors (LWR). The analytical portion of research project identified that a S-CO{sub 2} 'condensing' re-compression power cycle with multiple stages of reheat can increase LWR power conversion efficiency from 33-34% to 37-39%. The experimental portion of the project used Sandia's S-CO{sub 2} research loop to show that the as designed radial compressor could 'pump' liquid CO{sub 2} and that the gas-cooler's could 'condense' CO{sub 2} even though both of these S-CO{sub 2} components were designed to operate on vapor phase S-CO{sub 2} near the critical point. There is potentially very high value to this research as it opens the possibility of increasing LWR power cycle efficiency, above the 33-34% range, while lowering the capital cost of the power plant because of the small size of the S-CO{sub 2} power system. In addition it provides a way to incrementally build advanced LWRs that are optimally designed to couple to S-CO{sub 2} power conversion systems to increase the power cycle efficiency to near 40%.

In this work, we describe a novel design for a H2SO 4decomposer. The decomposition of H2SO4 to produce SO2is a common processing operation in the sulfur-based thermochemical cycles for hydrogen production where acid decomposition takes place at 850°C in the presence of a catalyst. The combination of high temperature and sulfuric acid creates a very corrosive environment that presents significant design challenges. The new decomposer design is based on a bayonet-type heat exchanger tube with the annular space packed with a catalyst. The unit is constructed of silicon carbide and other highly corrosion resistant materials. The new design integrates acid boiling, superheating, decomposition and heat recuperation into a single process and eliminates problems of corrosion and failure of high temperature seals encountered in previous testing using metallic construction materials. The unit was tested by varying the acid feed rate and decomposition temperature and pressure.

A new high-fidelity integrated system method and analysis approach was developed and implemented for consistent and comprehensive evaluations of advanced fuel cycles leading to minimized Transuranic (TRU) inventories. The method has been implemented in a developed code system integrating capabilities of Monte Carlo N - Particle Extended (MCNPX) for high-fidelity fuel cycle component simulations. In this report, a Nuclear Energy System (NES) configuration was developed to take advantage of used fuel recycling and transmutation capabilities in waste management scenarios leading to minimized TRU waste inventories, long-term activities, and radiotoxicities. The reactor systems and fuel cycle components that make up the NES were selected for their ability to perform in tandem to produce clean, safe, and dependable energy in an environmentally conscious manner. The diversity in performance and spectral characteristics were used to enhance TRU waste elimination while efficiently utilizing uranium resources and providing an abundant energy source. A computational modeling approach was developed for integrating the individual models of the NES. A general approach was utilized allowing for the Integrated System Model (ISM) to be modified in order to provide simulation for other systems with similar attributes. By utilizing this approach, the ISM is capable of performing system evaluations under many different design parameter options. Additionally, the predictive capabilities of the ISM and its computational time efficiency allow for system sensitivity/uncertainty analysis and the implementation of optimization techniques.

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.

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The impact associated with energy generation and utilization is immeasurable due to the immense, widespread, and myriad effects it has on the world and its inhabitants. The polar extremes are demonstrated on the one hand, by the high quality of life enjoyed by individuals with access to abundant reliable energy sources, and on the other hand by the global-scale environmental degradation attributed to the affects of energy production and use. Thus, nations strive to increase their energy generation, but are faced with the challenge of doing so with a minimal impact on the environment and in a manner that is self-reliant. Consequently, a revival of interest in nuclear energy has followed, with much focus placed on technologies for transmuting nuclear spent fuel. The performed research investigates nuclear energy systems that optimize the destruction of nuclear waste. In the context of this effort, nuclear energy system is defined as a configuration of nuclear reactors and corresponding fuel cycle components. The proposed system has unique characteristics that set it apart from other systems. Most notably the dedicated High-Energy External Source Transmuter (HEST), which is envisioned as an advanced incinerator used in combination with thermal reactors. The system is configured for examining environmentally benign fuel cycle options by focusing on minimization or elimination of high level waste inventories. Detailed high-fidelity exact-geometry models were developed for representative reactor configurations. They were used in preliminary calculations with Monte Carlo N-Particle eXtented (MCNPX) and Standardized Computer Analysis for Licensing Evaluation (SCALE) code systems. The reactor models have been benchmarked against existing experimental data and design data. Simulink{reg_sign}, an extension of MATLAB{reg_sign}, is envisioned as the interface environment for constructing the nuclear energy system model by linking the individual reactor and fuel component sub-models for overall analysis of the system. It also provides control over key user input parameters and the ability to effectively consolidate vital output results for uncertainty/sensitivity analysis and optimization procedures. The preliminary analysis has shown promising advanced fuel cycle scenarios that include Pressure Water Reactors Pressurized Water Reactors (PWRs), Very High Temperature Reactors (VHTRs) and dedicated HEST waste incineration facilities. If deployed, these scenarios may substantially reduce nuclear waste inventories approaching environmentally benign nuclear energy system characteristics. Additionally, a spent fuel database of the isotopic compositions for multiple design and control parameters has been created for the VHTR-HEST input fuel streams. Computational approaches, analysis metrics, and benchmark strategies have been established for future detailed studies.

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Energy demand and GDP per capita are strongly correlated, while public concern over the role of energy in climate change is growing. Nuclear power plants produce 16% of world electricity demands without greenhouse gases. Generation-IV advanced nuclear energy systems are being designed to be safe and economical. Minimizing the handling and storage of nuclear waste is important. NIF and ITER are bringing sustainable fusion energy closer, but a significant gap in fusion technology development remains. Fusion-fission hybrids could be a synergistic step to a pure fusion economy and act as a technology bridge. We discuss how a pulsed power-driven Z-pinch hybrid system producing only 20 MW of fusion yield can drive a sub-critical transuranic blanket that transmutes 1280 kg of actinide wastes per year and produces 3000 MW. These results are applicable to other inertial and magnetic fusion energy systems. A hybrid system could be introduced somewhat sooner because of the modest fusion yield requirements and can provide both a safe alternative to fast reactors for nuclear waste transmutation and a maturation path for fusion technology. The development and demonstration of advanced materials that withstand high-temperature, high-irradiation environments is a fundamental technology issue that is common to both fusion-fission hybrids and Generation-IV reactors. © 2008 Elsevier B.V. All rights reserved.

This report evaluates transportation risk for nuclear material in the proposed Global Nuclear Energy Partnership (GNEP) fuel cycle. Since many details of the GNEP program are yet to be determined, this document is intended only to identify general issues. The existing regulatory environment is determined to be largely prepared to incorporate the changes that the GNEP program will introduce. Nuclear material vulnerability and attractiveness are considered with respect to the various transport stages within the GNEP fuel cycle. It is determined that increased transportation security will be required for the GNEP fuel cycle, particularly for international transport. Finally, transportation considerations for several fuel cycle scenarios are discussed. These scenarios compare the current "once-through" fuel cycle with various aspects of the proposed GNEP fuel cycle.

The advent of the nuclear renaissance gives rise to a concern for the effective design of nuclear fuel cycle systems that are safe, secure, nonproliferating and cost-effective. We propose to integrate the monitoring of the four major factors of nuclear facilities by focusing on the interactions between Safeguards, Operations, Security, and Safety (SOSS). We proposed to develop a framework that monitors process information continuously and can demonstrate the ability to enhance safety, operations, security, and safeguards by measuring and reducing relevant SOSS risks, thus ensuring the safe and legitimate use of the nuclear fuel cycle facility. A real-time comparison between expected and observed operations provides the foundation for the calculation of SOSS risk. The automation of new nuclear facilities requiring minimal manual operation provides an opportunity to utilize the abundance of process information for monitoring SOSS risk. A framework that monitors process information continuously can lead to greater transparency of nuclear fuel cycle activities and can demonstrate the ability to enhance the safety, operations, security and safeguards associated with the functioning of the nuclear fuel cycle facility. Sandia National Laboratories (SNL) has developed a risk algorithm for safeguards and is in the process of demonstrating the ability to monitor operational signals in real-time though a cooperative research project with the Japan Atomic Energy Agency (JAEA). The risk algorithms for safety, operations and security are under development. The next stage of this work will be to integrate the four algorithms into a single framework.

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The advent of the nuclear renaissance gives rise to a concern for the effective design of nuclear fuel cycle systems that are safe, secure, nonproliferating and cost-effective. We propose to integrate the monitoring of the four major factors of nuclear facilities by focusing on the interactions between Safeguards, Operations, Security, and Safety (SOSS). We proposed to develop a framework that monitors process information continuously and can demonstrate the ability to enhance safety, operations, security, and safeguards by measuring and reducing relevant SOSS risks, thus ensuring the safe and legitimate use of the nuclear fuel cycle facility. A real-time comparison between expected and observed operations provides the foundation for the calculation of SOSS risk. The automation of new nuclear facilities requiring minimal manual operation provides an opportunity to utilize the abundance of process information for monitoring SOSS risk. A framework that monitors process information continuously can lead to greater transparency of nuclear fuel cycle activities and can demonstrate the ability to enhance the safety, operations, security and safeguards associated with the functioning of the nuclear fuel cycle facility. Sandia National Laboratories (SNL) has developed a risk algorithm for safeguards and is in the process of demonstrating the ability to monitor operational signals in real-time though a cooperative research project with the Japan Atomic Energy Agency (JAEA). The risk algorithms for safety, operations and security are under development. The next stage of this work will be to integrate the four algorithms into a single framework.

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This report evaluates transportation issues for nuclear material in the proposed Global Nuclear Energy Partnership (GNEP) fuel cycle. Since many details of the GNEP program are yet to be determined, this document is intended only to identify general issues. The existing regulatory environment is determined to be largely prepared to incorporate the changes that the GNEP program will introduce. Nuclear material vulnerability and attractiveness are considered with respect to the various transport stages within the GNEP fuel cycle. Physical protection options are then outlined for the transportation of this nuclear material. It is determined that increased transportation security will be required for the GNEP fuel cycle, particularly for international transport. Finally, transportation considerations for several fuel cycle scenarios are discussed. These scenarios compare the current "once-through" fuel cycle with various aspects of the proposed GNEP fuel cycle. 3

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We have developed a system of differential-output monitors that diagnose current and voltage in the vacuum section of a 20-MA 3-MV pulsed-power accelerator. The system includes 62 gauges: 3 current and 6 voltage monitors that are fielded on each of the accelerator's 4 vacuum-insulator stacks, 6 current monitors on each of the accelerator's 4 outer magnetically insulated transmission lines (MITLs), and 2 current monitors on the accelerator's inner MITL. The inner-MITL monitors are located 6 cm from the axis of the load. Each of the stack and outer-MITL current monitors comprises two separate B-dot sensors, each of which consists of four 3-mm-diameter wire loops wound in series. The two sensors are separately located within adjacent cavities machined out of a single piece of copper. The high electrical conductivity of copper minimizes penetration of magnetic flux into the cavity walls, which minimizes changes in the sensitivity of the sensors on the 100-ns time scale of the accelerator's power pulse. A model of flux penetration has been developed and is used to correct (to first order) the B-dot signals for the penetration that does occur. The two sensors are designed to produce signals with opposite polarities; hence, each current monitor may be regarded as a single detector with differential outputs. Common-mode-noise rejection is achieved by combining these signals in a 50-{Omega} balun. The signal cables that connect the B-dot monitors to the balun are chosen to provide reasonable bandwidth and acceptable levels of Compton drive in the bremsstrahlung field of the accelerator. A single 50-{omega} cable transmits the output signal of each balun to a double-wall screen room, where the signals are attenuated, digitized (0.5-ns/sample), numerically compensated for cable losses, and numerically integrated. By contrast, each inner-MITL current monitor contains only a single B-dot sensor. These monitors are fielded in opposite-polarity pairs. The two signals from a pair are not combined in a balun; they are instead numerically processed for common-mode-noise rejection after digitization. All the current monitors are calibrated on a 76-cm-diameter axisymmetric radial transmission line that is driven by a 10-kA current pulse. The reference current is measured by a current-viewing resistor (CVR). The stack voltage monitors are also differential-output gauges, consisting of one 1.8-cm-diameter D-dot sensor and one null sensor. Hence, each voltage monitor is also a differential detector with two output signals, processed as described above. The voltage monitors are calibrated in situ at 1.5 MV on dedicated accelerator shots with a short-circuit load. Faraday's law of induction is used to generate the reference voltage: currents are obtained from calibrated outer-MITL B-dot monitors, and inductances from the system geometry. In this way, both current and voltage measurements are traceable to a single CVR. Dependable and consistent measurements are thus obtained with this system of calibrated diagnostics. On accelerator shots that deliver 22 MA to a low-impedance z-pinch load, the peak lineal current densities at the stack, outer-MITL, and inner-MITL monitor locations are 0.5, 1, and 58 MA/m, respectively. On such shots the peak currents measured at these three locations agree to within 1%.

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The resurgence of interest in reprocessing in the United States with the Global Nuclear Energy Partnership has led to a renewed look at technologies for transmuting nuclear waste. Sandia National Laboratories has been investigating the use of a Z-Pinch fusion driver to burn actinide waste in a sub-critical reactor. The baseline design has been modified to solve some of the engineering issues that were identified in the first year of work, including neutron damage and fuel heating. An on-line control feature was added to the reactor to maintain a constant neutron multiplication with time. The transmutation modeling effort has been optimized to produce more accurate results. In addition, more attention was focused on the integration of this burner option within the fuel cycle including an investigation of overall costs. This report presents the updated reactor design, which is able to burn 1320 kg of actinides per year while producing 3,000 MWth.

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Robust and reliable quantitative proliferation assessment tools have the potential to contribute significantly to a strengthened nonproliferation regime and to the future deployment of nuclear fuel cycle technologies. Efforts to quantify proliferation resistance have thus far met with limited success due to the inherent subjectivity of the problem and interdependencies between attributes that lead to proliferation resistance. We suggest that these limitations flow substantially from weaknesses in the foundations of existing methodologies--the initial data inputs. In most existing methodologies, little consideration has been given to the utilization of varying types of inputs--particularly the mixing of subjective and objective data--or to identifying, understanding, and untangling relationships and dependencies between inputs. To address these concerns, a model set of inputs is suggested that could potentially be employed in multiple approaches. We present an input classification scheme and the initial results of testing for relationships between these inputs. We will discuss how classifying and testing the relationship between these inputs can help strengthen tools to assess the proliferation risk of nuclear fuel cycle processes, systems, and facilities.

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Proliferation resistance features that reduce the likelihood of diversion of nuclear materials from the civilian nuclear power fuel cycle are critical for a global nuclear future. A framework that monitors process information continuously can demonstrate the ability to resist proliferation by measuring and reducing diversion risk, thus ensuring the legitimate use of the nuclear fuel cycle. The automation of new nuclear facilities requiring minimal manual operation makes this possible by generating instantaneous system state data that can be used to track and measure the status of the process and material at any given time. Sandia National Laboratories (SNL) and the Japan Atomic Energy Agency (JAEA) are working in cooperation to develop an advanced transparency framework capable of assessing diversion risk in support of overall plant transparency. The ''diversion risk'' quantifies the probability and consequence of a host nation diverting nuclear materials from a civilian fuel cycle facility. This document introduces the details of the diversion risk quantification approach to be demonstrated in the fuel handling training model of the MONJU Fast Reactor.

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Due to increasing concerns over the buildup of long-lived transuranic isotopes in spent nuclear fuel waste, attention has been given in recent years to technologies that can burn up these species. The separation and transmutation of transuranics is part of a solution to decreasing the volume and heat load of nuclear waste significantly to increase the repository capacity. A fusion neutron source can be used for transmutation as an alternative to fast reactor systems. Sandia National Laboratories is investigating the use of a Z-Pinch fusion driver for this application. This report summarizes the initial design and engineering issues of this ''In-Zinerator'' concept. Relatively modest fusion requirements on the order of 20 MW can be used to drive a sub-critical, actinide-bearing, fluid blanket. The fluid fuel eliminates the need for expensive fuel fabrication and allows for continuous refueling and removal of fission products. This reactor has the capability of burning up 1,280 kg of actinides per year while at the same time producing 3,000 MWth. The report discusses the baseline design, engineering issues, modeling results, safety issues, and fuel cycle impact.

This report summarizes the work conducted for the Z-inertial fusion energy (Z-IFE) late start Laboratory Directed Research Project. A major area of focus was on creating a roadmap to a z-pinch driven fusion power plant. The roadmap ties ZIFE into the Global Nuclear Energy Partnership (GNEP) initiative through the use of high energy fusion neutrons to burn the actinides of spent fuel waste. Transmutation presents a near term use for Z-IFE technology and will aid in paving the path to fusion energy. The work this year continued to develop the science and engineering needed to support the Z-IFE roadmap. This included plant system and driver cost estimates, recyclable transmission line studies, flibe characterization, reaction chamber design, and shock mitigation techniques.

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A key objective to the global deployment of nuclear technology is maintaining transparency among nation-states and international communities. By providing an environment in which to exchange scientific and technological information regarding nuclear technology, the safe and legitimate use of nuclear material and technology can be assured. Many nations are considering closed or multiple-application nuclear fuel cycles and are subsequently developing advanced reactors in an effort to obtain some degree of energy self-sufficiency. Proliferation resistance features that prevent theft or diversion of nuclear material and reduce the likelihood of diversion from the civilian nuclear power fuel cycle are critical for a global nuclear future. IAEA Safeguards have been effective in minimizing opportunities for diversion; however, recent changes in the global political climate suggest implementation of additional technology and methods to ensure the prompt detection of proliferation. For a variety of reasons, nuclear facilities are becoming increasingly automated and will require minimum manual operation. This trend provides an opportunity to utilize the abundance of process information for monitoring proliferation risk, especially in future facilities. A framework that monitors process information continuously can lead to greater transparency of nuclear fuel cycle activities and can demonstrate the ability to resist proliferation associated with these activities. Additionally, a framework designed to monitor processes will ensure the legitimate use of nuclear material. This report describes recent efforts to develop a methodology capable of assessing proliferation risk in support of overall plant transparency. The framework may be tested at the candidate site located in Japan: the Fuel Handling Training Model designed for the Monju Fast Reactor at the International Cooperation and Development Training Center of the Japan Atomic Energy Agency.

On a single-pulse basis, the tungsten armor for the chamber walls in a laser inertial fusion energy power plant must withstand X-ray fluences of 0.4-1.2 J/cm2 with almost no mass loss, and preferably no surface changes. We have exposed preheated tungsten samples to 0.27 and 0.9 J/cm 2 X-ray fluence from the Z accelerator at Sandia National Laboratories to determine the single-shot X-ray damage threshold. Earlier focused ion beam analysis has shown that rolled powdered metal formed tungsten and tungsten alloys, will melt when exposed to 2.3 J/cm2 on Z, but not at 1.3 J/cm2. Three forms of tungsten - single-crystal (SING), chemical-vapor-deposited (CVD), and rolled powdered metal (PWM) - were exposed to fluence levels of 0.9 J/cm2 without any apparent melting. However, the CVD and PWM sample surfaces were rougher after exposure than the SING sample, which was not roughened. BUCKY (1D) calculations show a threshold of 0.5 J/cm2 for melting on Z. The present experiments indicate no melting but limited surface changes occur with polycrystalline samples (PWM and CVD) at 0.9 J/cm2 and no surface changes other than debris for samples at 0.27 J/cm2. © 2005 Elsevier B.V. All rights reserved.

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Nuclear fuel cycle transparency can be defined as a confidence building approach among political entities to ensure civilian nuclear facilities are not being used for the development of nuclear weapons. Transparency concepts facilitate the transfer of nuclear technology, as the current international political climate indicates a need for increased methods of assuring non-proliferation. This research develops a system which will augment current non-proliferation assessment activities undertaken by U.S. and international regulatory agencies. It will support the export of nuclear technologies, as well as the design and construction of Gen. IV energy systems. Additionally, the framework developed by this research will provide feedback to cooperating parties, thus ensuring full transparency of a nuclear fuel cycle. As fuel handling activities become increasingly automated, proliferation or diversion potential of nuclear material still needs to be assessed. However, with increased automation, there exists a vast amount of process data to be monitored. By designing a system that monitors process data continuously, and compares this data to declared process information and plant designs, a faster and more efficient assessment of proliferation risk can be made. Figure 1 provides an illustration of the transparency framework that has been developed. As shown in the figure, real-time process data is collected at the fuel cycle facility; a reactor, a fabrication plant, or a recycle facility, etc. Data is sent to the monitoring organization and is assessed for proliferation risk. Analysis and recommendations are made to cooperating parties, and feedback is provided to the facility. The analysis of proliferation risk is based on the following factors: (1) Material attractiveness: the quantification of factors relevant to the proliferation risk of a certain material (e.g., highly enriched Pu-239 is more attractive than that of lower enrichment) (2) The static (baseline) risk: the quantification of risk factors regarding the expected value of proliferation risk under normal (not proliferating) operations. (3) The dynamic (changing) risk: the quantification of risk factors regarding the observed value of proliferation risk, based on monitor signals from facility operations. This framework could be implemented at facilities which have been exported (for instance, to third world countries), or facilities located in sensitive countries. Sandia National Laboratories is currently working with the Japan Nuclear Cycle Development Institute (JNC) to implement a demonstration of nuclear fuel cycle transparency technology at the Fuel Handling Training Model designed for the Monju Fast Reactor at the International Cooperation and Development Training Center in Japan. This technology has broad applications, both in the U.S. and abroad. Following the demonstration, we expect to begin further testing of the technology at an Enrichment Facility, a Fast Reactor, and at a Recycle Facility.

Concerns over the illicit trafficking of radiological and nuclear materials were focused originally on the lack of security and accountability of such material throughout the former Soviet states. This is primarily attributed to the frequency of events that have occurred involving the theft and trafficking of critical material components that could be used to construct a Radiological Dispersal Device (RDD) or even a rudimentary nuclear device. However, with the continued expansion of nuclear technology and the deployment of a global nuclear fuel cycle these materials have become increasingly prevalent, affording a more diverse inventory of dangerous materials and dual-use items. To further complicate the matter, the list of nuclear consumers has grown to include: (1) Nation-states that have gone beyond the IAEA agreed framework and additional protocols concerning multiple nuclear fuel cycles and processes that reuse the fuel through reprocessing to exploit technologies previously confined to the more industrialized world; (2) Terrorist organizations seeking to acquire nuclear and radiological material due to the potential devastation and psychological effect of their use; (3) Organized crime, which has discovered a lucrative market in trafficking of illicit material to international actors and/or countries; and (4) Amateur smugglers trying to feed their families in a post-Soviet era. An initial look at trafficking trends of this type seems scattered and erratic, localized primarily to a select group of countries. This is not necessarily the case. The success with which other contraband has been smuggled throughout the world suggests that nuclear trafficking may be carried out with relative ease along the same routes by the same criminals or criminal organizations. Because of the inordinately high threat posed by terrorist or extremist groups acquiring the ingredients for unconventional weapons, it is necessary that illicit trafficking of these materials be better understood as to prepare for the sustained global development of the nuclear fuel cycle. Conversely, modeling and analyses of this activity must not be limited in their scope to loosely organized criminal smuggling, but address the problem as a commercial, industrial project for the covert development of nuclear technologies and unconventional weapon development.

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The Z-Pinch Power Plant uses the results from Sandia National Laboratories Z accelerator in a power plant application to generate energy pulses using inertial confinement fusion. A collaborative project has been initiated by Sandia to investigate the scientific principles of a power generation system using this technology. Research is under way to develop an integrated concept that describes the operational issues of a 1000 MW electrical power plant. Issues under consideration include: 1-20 gigajoule fusion pulse containment, repetitive mechanical connection of heavy hardware, generation of terawatt pulses every 10 seconds, recycling of ten thousand tons of steel, and manufacturing of millions of hohlraums and capsules per year. Additionally, waste generation and disposal issues are being examined. This paper describes the current concept for the plant and also the objectives for future research.

A two-year effort focused on applying ASCI technology developed for the analysis of weapons systems to the state-of-the-art accident analysis of a nuclear reactor system was proposed. The Sandia SIERRA parallel computing platform for ASCI codes includes high-fidelity thermal, fluids, and structural codes whose coupling through SIERRA can be specifically tailored to the particular problem at hand to analyze complex multiphysics problems. Presently, however, the suite lacks several physics modules unique to the analysis of nuclear reactors. The NRC MELCOR code, not presently part of SIERRA, was developed to analyze severe accidents in present-technology reactor systems. We attempted to: (1) evaluate the SIERRA code suite for its current applicability to the analysis of next generation nuclear reactors, and the feasibility of implementing MELCOR models into the SIERRA suite, (2) examine the possibility of augmenting ASCI codes or alternatives by coupling to the MELCOR code, or portions thereof, to address physics particular to nuclear reactor issues, especially those facing next generation reactor designs, and (3) apply the coupled code set to a demonstration problem involving a nuclear reactor system. We were successful in completing the first two in sufficient detail to determine that an extensive demonstration problem was not feasible at this time. In the future, completion of this research would demonstrate the feasibility of performing high fidelity and rapid analyses of safety and design issues needed to support the development of next generation power reactor systems.

A major cause of failures in heat exchangers and steam generators in nuclear power plants is degradation of the tubes within them. The tube failure is often caused by the development of cracks that begin on the outer surface of the tube and propagate both inwards and laterally. A new technique was researched for detection of defects using a continuous-wave radar method within metal tubing. The experimental program resulted in a completed product development schedule and the design of an experimental apparatus for studying handling of the probe and data acquisition. These tests were completed as far as the prototypical probe performance allowed. The prototype probe design did not have sufficient sensitivity to detect a defect signal using the defined radar technique and did not allow successful completion of all of the project milestones. The best results from the prototype probe could not detect a tube defect using the radar principle. Though a more precision probe may be possible, the cost of design and construction was beyond the scope of the project. This report describes the probe development and the status of the design at the termination of the project.

Abstract not provided.

With the promising new results of fast z-pinch technology developed at Sandia National Laboratories, we are investigating using z-pinch driven high-yield Inertial Confinement Fusion (ICF) as a fusion power plant energy source. These investigations have led to a novel fusion system concept based on an attempt to separate many of the difficult fusion engineering issues and a strict reliance on existing technology, or a reasonable extrapolation of existing technology, wherever possible. In this paper, we describe the main components of such a system with a focus on the fusion chamber dynamics. The concept works with all of the electrically-coupled ICF proposed fusion designs. It is proposed that a z-pinch driven ICF power system can be feasibly operated at high yields (1 to 30 GJ) with a relatively low pulse rate (0.01-0.1 Hz). To deliver the required current from the rep-rated pulse power driver to the z-pinch diode, a Recyclable Transmission Line (RTL) and the integrated target hardware are fabricated, vacuum pumped, and aligned prior to loading for each power pulse. In this z-pinch driven system, no laser or ion beams propagate in the chamber such that the portion of the chamber outside the RTL does not need to be under vacuum. Additionally, by utilizing a graded-density solid lithium or fluorine/lithium/beryllium eutectic (FLiBe) blanket between the source and the first-wall the system can breed its own fuel absorb a large majority of the fusion energy released from each capsule and shield the first-wall from a damaging neutron flux. This neutron shielding significantly reduces the neutron energy fluence at the first-wall such that radiation damage should be minimal and will not limit the first-wall lifetime. Assuming a 4 m radius, 8 m tall cylindrical chamber design with an 80 cm thick spherical FLiBe blanket, our calculations suggest that a 20 cm thick 6061-T6 Al chamber wall will reach the equivalent uranium ore radioactivity level within 100 years after a 30 year plant operation. The implication of this low radioactivity is that a z-pinch driven power plant may not require deep geologic waste storage.

The Z machine at Sandia National Laboratories (SNL) is the most powerful multi-module synchronized pulsed-power accelerator in the world. Rapid development of z-pinch loads on Z has led to outstanding progress in the last few years, resulting in radiative powers of up to 280 TW in 4 ns and a total radiated x-ray energy of 1.8 MJ. The present goal is to demonstrate single-shot, high-yield fusion capsules. Pulsed power is a robust and inexpensive technology, which should be well suited for Inertial Fusion Energy, but a rep-rated capability is needed. Recent developments have led to a viable conceptual approach for a rep-rated z-pinch power plant for IFE. This concept exploits the advantages of going to high yield (a few GJ) at low rep-rate ({approximately} 0.1 Hz), and using a Recyclable Transmission Line (RTL) to provide the necessary standoff between the fusion target and the power plant chamber. In this approach, a portion of the transmission line near the capsule is replaced after each shot. The RTL should be constructed of materials that can easily be separated from the liquid coolant stream and refabricated for a subsequent shots. One possibility is that most of the RTL is formed by casting FLiBe, a salt composed of fluorine, lithium, and beryllium, which is an attractive choice for the reactor coolant, with chemically compatible lead or tin on the surface to provide conductivity. The authors estimate that fusion yields greater than 1 GJ will be required for efficient generation of electricity. Calculations indicate that the first wall will have an acceptable lifetime with these high yields if blast mitigation techniques are used. Furthermore, yields above 5 GJ may allow the use of a compact blanket direct conversion scheme.

In principle, the energy released by a fission can be converted directly into electricity by using the charged fission fragments. The first theoretical treatment of direct energy conversion (DEC) appeared in the literature in 1957. Experiments were conducted over the next ten years, which identified a number of problem areas. Research declined by the late 1960's due to technical challenges that limited performance. Under the Nuclear Energy Research Initiative the authors are determining if these technical challenges can be overcome with todays technology. The authors present the basic principles of DEC reactors, review previous research, discuss problem areas in detail, and identify technological developments of the last 30 years that can overcome these obstacles. As an example, the fission electric cell must be insulated to avoid electrons crossing the cell. This insulation could be provided by a magnetic field as attempted in the early experiments. However, from work on magnetically insulated ion diodes they know how to significantly improve the field geometry. Finally, a prognosis for future development of DEC reactors will be presented .

In Inertial Fusion Energy (IFE), Target Chamber Dynamics (TCD) is an integral part of the target chamber design and performance. TCD includes target output deposition of target x-rays, ions and neutrons in target chamber gases and structures, vaporization and melting of target chamber materials, radiation-hydrodynamics in target chamber vapors and gases, and chamber conditions at the time of target and beam injections. Pulsed power provides a unique environment for IFE-TCD validation experiments in two important ways: they do not require the very clean conditions which lasers need and they currently provide large x-ray and ion energies.