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All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

The Department of Energy's (DOE's) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

Smith, Cyrus M.↗
DOE OSTI2023

AGR-5/6/7 Irradiation Disassembly and Metrology First Look

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform research and development on tristructural isotropic (TRISO)-coated particle fuel to support deployment of high-temperature gas-cooled reactors (HTGRs), which are graphite-moderated nuclear reactors cooled with helium. This work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel Program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing, deployment, and operation of HTGRs in the United States. To achieve these goals, the program includes fuel fabrication, irradiations of TRISO fuels and high-temperature materials (e.g., graphite), safety testing and post-irradiation examination (PIE), fuel performance modeling, and fission product transport and source term determination. The ART AGR program has conducted four distinct fuel irradiation experiments in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The first of these irradiation tests, designated AGR-1, began in ATR in December of 2006 and ended in November 2009. This experiment was primarily to act as a shakedown test of the multi-capsule test train design and to provide early data on fuel performance that would be used in fuel fabrication process development. AGR-1 fuel kernels were produced on an engineering scale, but the TRISO coatings and cylindrical fuel compacts were fabricated on a laboratory scale. The AGR-1 PIE was completed and the final report was published in 2015. The second irradiation test, AGR-2, started in ATR in June 2010 and ended in October 2013. The AGR-2 irradiation test was designed to provide fuel performance data for coated particles fabricated on an engineering-scale pilot line using a coater with an internal chamber diameter of 150 mm (6 in.). The final PIE report was published in 2021. AGR-3/4, a single irradiation that combined what were originally conceived as the third and fourth tests, was to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport. PIE of the AGR-3/4 experiment is still in progress as of this writing. The subject of this report is AGR-5/6/7, the final qualification test of AGR TRISO fuel made entirely at the engineering scale.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

A Proposal for a Standard Interface Between Monte Carlo Tools And One-Loop Programs

Many highly developed Monte Carlo tools for the evaluation of cross sections based on tree matrix elements exist and are used by experimental collaborations in high energy physics. As the evaluation of one-loop matrix elements has recently been undergoing enormous progress, the combination of one-loop matrix elements with existing Monte Carlo tools is on the horizon. This would lead to phenomenological predictions at the next-to-leading order level. This note summarises the discussion of the next-to-leading order multi-leg (NLM) working group on this issue which has been taking place during the workshop on Physics at TeV Colliders at Les Houches, France, in June 2009. The result is a proposal for a standard interface between Monte Carlo tools and one-loop matrix element programs.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2022

Consensus DOE Advanced Fuels Campaign TREAT/SATS Test Plan [Slides]

This record is comprised of summary slides of the Combined TREAT-LOC & SATS Integral LOCA Experiment Plan. The experimental program has been developed to specifically address data gaps and opportunities identified through detailed review of the existing public knowledgebase on LOCA FFRD and specific experimental development for prototypic LOCA conditions for LWR systems. The test program relies on a unique combination of in-pile and out-of-pile experimental approaches to (1) provide clear tieback to the existing integral and semi-integral LOCA experiment database using state-of-the-art facilities. More importantly, this program will systematically investigate the impacts of: (2) prototypic HBu fuel and cladding thermomechanical behaviors under postulated LWR LOCA conditions never fully investigated before. These conditions correspond with prototypic decay-energy heat up (DEH) and stored-energy heat up (SEH) conditions. Unique TREAT capability will provide first evaluation of SEH conditions on HBu fuels. The test program includes an emphasis on developing improved mechanistic understanding of key phenomena through independent experimental systems, development of a database to support fuel performance modeling tools, world leading advanced materials characterization, and the most advanced approach to in-situ diagnostics ever deployed to evaluate FFRD. The results will represent a significant leap forward in the evaluation of prototypic conditions and novel data to support modeling development and validation, as well as to inform the technical basis of LOCA-induced FFRD.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

A GPU Accelerated Simulation Program for Electron Cooling Process

Electron cooling is essential to achieve high luminosity in the medium energy electron ion collider (MIEC) project at Jefferson Lab. Bunched electron beam with energy above 50 MeV is used to cool coasting and/or bunched ion beams. Although the conventional electron cooling technique has been widely used, such an implementation in MEIC is very challenging. We are developing a simulation program for the electron cooling process to fulfill the need of the electron cooling system design for MEIC. The program simulates the evolution of the ion beam under the intrabeam scattering (IBS) effect and the electron cooling effect using Monte Carlo method. To accelerate the calculation, the program is developed on a GPU platform. We will present some preliminary simulation results.

Zhang, He↗
DOE OSTI2012

Initial Experiences with the NEAMS Early User Program

In FY 2012, the NEAMS program initiated an Early User Program to get the modeling and simulation codes being developed by the program into the hands of “friendly” users outside the development team. The primary purpose of this activity was to provide the development teams with independent feedback as to the scientific and computational performance of the codes, their usability and other factors. It also serves to provide the stakeholders of the NEAMS program with visibility into the progress and capability of the codes prior to their anticipated first full official release in 2018. Seven teams were supported under the FY 2012 Early User program, working with the BISON/MOOSE and AMP fuel performance codes, and the Nek5000 computational fluid dynamics code for reactor modeling, primarily in validation studies. The program was considered successful and useful by the Product Line Leads, and worked largely as expected. However we have noted a number of areas in which the program could be adjusted and improved in the future. The most significant have to do with ensuring Early Users get training on the applications as early as possible, and finding ways to reduce some of the constraints imposed by the NEAMS financial schedule and the effects of Continuing Resolutions.

Baird, Mark L.↗
DOE OSTI2015

Developing MiniApps on Modern Platforms Using Multiple Programming Models, In: 2015 IEEE International Conference on Cluster Computing

We have developed a set of reduced, proxy applications ("MiniApps") based on large-scale application codes supported at the Oak Ridge Leadership Computing Facility (OLCF). The MiniApps are designed to encapsulate the details of the most important (i.e. the most time-consuming and/or unique) facets of the applications that run in production mode on the OLCF. In each case, we have produced or plan to produce individual versions of the MiniApps using different specific programming models (e.g., OpenACC, CUDA, OpenMP). We describe some of our initial observations regarding these different implementations along with estimates of how closely the MiniApps track the actual performance characteristics (in particular, the overall scalability) of the large-scale applications from which they are derived.

Computer Science↗
DOE OSTI2023

Advanced Simulation and Computing: ASC FY24 Implementation Plan

The DOE National Nuclear Security Administration (NNSA) Stockpile Stewardship Program (SSP) is an integrated technical program for maintaining the safety, security, and reliability of the U.S. nuclear stockpile. The SSP incorporates nuclear test data, computational modeling and simulation, and experimental facilities to advance understanding of nuclear weapons. The suite of data analyzed comes from activities including previous nuclear tests, stockpile surveillance, experimental research, and development and engineering programs. This integrated national program requires the continued use of experimental facilities and the computational capabilities to support the SSP missions. These component parts, in addition to an appropriately scaled production capability, enable NNSA to support stockpile requirements. The ultimate goal of the SSP, and thus of the Advanced Simulation and Computing (ASC) program, is to ensure that the U.S. maintains a safe, secure, and effective strategic deterrent. The ASC program is a cornerstone of the SSP, providing simulation capabilities and computational resources to support the annual stockpile assessment and certification process, study advanced nuclear weapons design and manufacturing processes, analyze accident scenarios and weapons aging, and provide the tools to enable stockpile Life Extension Programs (LEPs) and the resolution of Significant Finding Investigations (SFIs). This work requires a balance of resources, including technical staff, hardware, simulation software, and computer science solutions. The ASC program focuses on increasing the predictive capabilities in a three-dimensional (3D) simulation environment while maintaining support to the SSP. The Program continues to improve its unique tools for understanding and solving progressively more difficult stockpile problems (sufficient resolution, dimensionality, and scientific details), and quantifying critical margins and uncertainties. Resolving each issue requires increasingly difficult analyses because the aging process has progressively moved the stockpile further from the original test base. While the focus remains on the U.S. nuclear weapons program, where possible, the Program also enables the use of high-performance computing (HPC) and simulation tools to address broader national security needs, such as foreign nuclear weapon assessments and nuclear counterterrorism. The 2022 Nuclear Posture Review (NPR) calls for NNSA to “deliver a modern, adaptive nuclear security enterprise based on an integrated strategy for risk management, production-based resilience, science and technology innovation, and workforce initiatives.” Furthermore, “NNSA will establish a Science and Technology Innovation Initiative to accelerate the integration of science and technology (S&T) throughout its activities.” Executing this strategy necessitates the continued emphasis on developing and sustaining high-quality scientific and engineering staff, as well as supporting computational and experimental capabilities. These components constitute the foundation of the nuclear weapons program. The continued success of the SSP and LEPs is predicated upon the ability to credibly certify the stockpile, without a return to underground nuclear tests (UGTs). Shortly after the nuclear test moratorium entered into force in 1992, the Accelerated Strategic Computing Initiative (ASCI) was established to provide an extensive simulation capability to underpin stockpile certification. While computing and simulation have always been essential to the success of the nuclear weapons program, the program goal of ASCI was to execute NNSA’s vision of using these tools in support of the stockpile stewardship mission. The ASCI program was essential to the successful demonstration of the SSP, providing critical nuclear weapons simulation and modeling capabilities. ASCI officially evolved into the ASC program in fiscal year (FY) 2005, but the mission remains essentially the same: provide the simulation and computational capabilities that underpin the ability to maintain a safe, secure, effective nuclear weapon stockpile, without returning to underground nuclear testing. The capabilities that the ASC program provides at the national laboratories play a vital role in the nuclear security enterprise and are necessary for fulfilling the stockpile stewardship and life extension requirements outlined for NNSA. The Program develops modern simulation tools that provide insights into stockpile aging issues, provide the computational and simulation tools that enable designers and analysts to certify the current stockpile and life-extended nuclear weapons, and inform the decision-making process when any modifications in nuclear warheads or the associated manufacturing processes are deemed necessary. Furthermore, ASC is enhancing the predictive simulation capabilities that are essential to evaluate weapons effects, design experiments, and ensure test readiness. The ASC program continues to improve its unique tools to solve stockpile problems— with a focus on sufficient resolution, dimensionality, and scientific detail—to enable Quantification of Margins and Uncertainties (QMU) and to resolve the increasingly difficult analyses needed for stockpile stewardship. The needs of the Stockpile Management and Production Modernization programs (formerly Directed Stockpile Work) also drive the requirements for simulation and computational resources. These requirements include planned LEPs, stockpile support activities, and mitigation efforts against the potential for technical surprise. All of the weapons within the current stockpile are in some stage of the life extension process. The simulation and computational capabilities are crucial for successful execution of these life extensions and for ensuring NNSA can certify these life-extended weapons without conducting a UGT.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Cycle-by-cycle Delay Estimation at Signalized Intersections by using Machine Learning and Simulated Video Detection Data

Accurate estimation of delay is crucial for efficient traffic signal operations. Estimation of delay in the real-time manner using traditional loop detectors requires advanced detectors (in addition to stop-bar detection). In cases when this detection layout is not in place, delay estimates are approximated with a lower accuracy. Video detection is one of the most frequently deployed detection systems at signalized intersections in recent years. In most cases video detection operates in the same way as traditional inductive loops. However, when coupled with computer vision algorithms, video detection systems could be used to retrieve additional information (e.g., vehicular arrivals and departures) that cannot be taken out from the conventional systems (e.g., long stop-bar loop detectors). Although present for several decades, video detection data were not frequently examined for delay estimation purposes. In this study, we proposed a novel delay estimation model which can be developed with only data from stop-bar video detectors. Relevant data were collected from a simulation model of 11 signalized intersections at downtown Chattanooga, TN and processed to create needed inputs for model development. With the use of multigene genetic programming the authors developed a delay model that outperforms accuracy of multi regression model. Furthermore, authors evaluated the developed model by comparison with the other benchmark delay models, such as HCM and approach delay model. It was found that the developed MGGP delay model outperforms benchmark models for a wide range of traffic and signal operation conditions.

Erdagi, Ismet Goksad↗
DOE OSTI2014

Measurements of 222 Rn, 220 Rn, and CO 2 Emissions in Natural CO 2 Fields in Wyoming: MVA Techniques for Determining Gas Transport and Caprock Integrity

An integrated field-laboratory program evaluated the use of radon and CO 2 flux measurements to constrain source and timescale of CO 2 fluxes in environments proximate to CO 2 storage reservoirs. By understanding the type and depth of the gas source, the integrity of a CO 2 storage reservoir can be assessed and monitored. The concept is based on correlations of radon and CO 2 fluxes observed in volcanic systems. This fundamental research is designed to advance the science of Monitoring, Verification, and Accounting (MVA) and to address the Carbon Storage Program goal of developing and validating technologies to ensure 99 percent storage performance. Graduate and undergraduate students conducted the research under the guidance of the Principal Investigators; in doing so they were provided with training opportunities in skills required for implementing and deploying CCS technologies. Although a final method or “tool” was not developed, significant progress was made. The field program identified issues with measuring radon in environments rich in CO 2 . Laboratory experiments determined a correction factor to apply to radon measurements made in CO 2 -bearing environments. The field program also identified issues with radon and CO 2 -flux measurements in soil gases at a natural CO 2 analog. A systematic survey of radon and CO 2 flux in soil gases at the LaBarge CO 2 Field in Southwest Wyoming indicates that measurements of 222 Rn (radon), 220 Rn (thoron), and CO 2 flux may not be a robust method for monitoring the integrity of a CO 2 storage reservoir. The field program was also not able to correlate radon and CO 2 flux in the CO 2 -charged springs of the Thermopolis hydrothermal system. However, this part of the program helped to motivate the aforementioned laboratory experiments that determined correction factors for measuring radon in CO 2 -rich environments. A graduate student earned a Master of Science degree for this part of the field program; she is currently employed with a geologic consulting company. Measurement of radon in springs has improved significantly since the field program first began; however, in situ measurement of 222 Rn and particularly 220 Rn in springs is problematic. Future refinements include simultaneous salinity measurements and systematic corrections, or adjustments to the partition coefficient as needed for more accurate radon concentration determination. A graduate student earned a Master of Science degree for this part of the field program; he is currently employed with a geologic consulting company. Both graduate students are poised to begin work in a CCS technology area. Laboratory experiments evaluated important process-level fundamentals that effect measurements of radon and CO 2 . Laboratory tests established that fine-grained source minerals yield higher radon emissivity compared to coarser-sized source minerals; subtleties in the dataset suggest that grain size alone is not fully representative of all the processes controlling the ability of radon to escape its mineral host. Emissivity for both 222 Rn and 220Rn increases linearly with temperature due to reaction of rocks with water, consistent with faster diffusion and enhanced mineral dissolution at higher temperatures. The presence of CO 2 changes the relative importance of the factors that control release of radon. Emissivity for both 222 Rn and 220 Rn in CO 2 -bearing experiments is greater at all temperatures compared to the experiments without CO 2 , but emissivity does not increase as a simple function of temperature. Governing processes may include a balance between enhanced dissolution versus carbonate mineral formation in CO 2 -rich waters.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Development of a Mine Rescue Drilling System (MRDS)

Sandia National Laboratories (Sandia) has a long history in developing compact, mobile, very high-speed drilling systems and this technology could be applied to increasing the rate at which boreholes are drilled during a mine accident response. The present study reviews current technical approaches, primarily based on technology developed under other programs, analyzes mine rescue specific requirements to develop a conceptual mine rescue drilling approach, and finally, proposes development of a phased mine rescue drilling system (MRDS) that accomplishes (1) development of rapid drilling MRDS equipment; (2) structuring improved web communication through the Mine Safety & Health Administration (MSHA) web site; (3) development of an improved protocol for employment of existing drilling technology in emergencies; (4) deployment of advanced technologies to complement mine rescue drilling operations during emergency events; and (5) preliminary discussion of potential future technology development of specialized MRDS equipment. This phased approach allows for rapid fielding of a basic system for improved rescue drilling, with the ability to improve the system over time at a reasonable cost.

42 ENGINEERING↗
DOE OSTI2012

Enterprise SRS: Leveraging Ongoing Operations to Advance Nuclear Fuel Cycle Programs - 12579

The international leadership in nuclear technology development and deployment long held by the United States has eroded due to the lack of clear national strategies for advanced reactor fuel cycle concepts and for nuclear materials management, as well as to the recent policy decision that halts work on the nuclear fuel repository at Yucca Mountain. Although no national consensus on strategy has yet been reached, a number of recent high-profile reviews and workshops have clearly highlighted a national need for robust research, development and deployment (RD and D) programs in key areas of nuclear technology, especially nuclear separations science and engineering. Collectively, these reviews and workshops provide a picture of the nuclear separations mission needs for three major program offices: Department of Energy Office of-Environmental Management), DOE Office of Nuclear Energy), and the National Nuclear Security Administration (NNSA). While the individual program needs differ significantly in detail and timing, they share common needs in two critical areas of RD and D: - The need for access to and use of multi-purpose engineering-scale demonstration test facilities that can support testing with radioactive material, and - The need for collaborative research enterprises that encompass government research organizations (i.e., national laboratories), commercial industry and the academic community. Such collaborative enterprises effectively integrate theory and modeling with the actual experimental work at all scales, as well as strengthen the technical foundation for research in critical areas. The arguments for engineering-scale collaborative research facilities are compelling. Processing history has shown that test programs and demonstrations conducted with actual nuclear materials are essential to program success. It is widely recognized, however, that such facilities are expensive to build and maintain; creating an imposing, if not prohibitive, financial burden on an individual sponsoring office. Given that reality, success for the current and future nuclear separations missions is dependent on a concerted effort to develop new, creative, approaches that leverage existing facilities in a manner that supports both near- and long-term needs of national programs. As a result of this situation, the Savannah River National Laboratory (SRNL) organized the 'Nuclear Separations User Facility Strategy Session' in Washington, D.C. on July 29, 2011. This workshop brought together key stakeholders from DOE and the private sector to develop a strategy for using engineering-scale nuclear materials processing facilities to advance our nation's nuclear separations research needs. In particular, the meeting focused on recommending how these engineering-scale demonstration facilities, like the Savannah River Site H-Canyon, can be connected with smaller 'bench-scale' research activities to form a seamless approach that integrates across the continuum of RD and D of advanced separations technologies. Coming out of this workshop, a new vision has been developed for a collaborative research facility model that centers on H-Canyon. Unique to this approach is the fact that H-Canyon will continue to accomplish DOE's critical nuclear material processing missions, while simultaneously serving as an RD and D resource for the scientific and technical portions of the nuclear separations community. This paper describes the planned operations for H-Canyon in FY2012 and beyond and discusses how these operations fit within the context of a collaborative research facility model and support the ongoing fuel cycle research and development programs of the DOE. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Light Water Reactor Sustainability Program — A Summary of Collaborative Research and Development Activities

This report describes the collaborative research and development projects in which the U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) Program is engaged with industry, regulatory agencies, and other organizations to achieve outcomes that are directed at sustaining the U.S. nuclear power industryduring Fiscal Year 2018. One way that this Program accomplishes its objectives—to develop technologies and other solutions that can improve the reliability, sustain the safety, enhance the economic performance, and extend the life of current reactors—is by conducting research together with industry and key organizations through cooperative arrangements and plans, through coordinated planning and management of resources and activities, through periodic technical exchanges of data and information, or by other means of interaction (i.e., contracts, memoranda of understanding, non-disclosure agreements, etc.). Broadly, these interactions represent different types of collaborations—formal or informal agreements between two or more organizations to work on something together that may be for a short or extended duration and may involve a single purpose or broad set of goals. DOE’s research, development, and demonstration role focuses on enhancing the safe, efficient, and economical performance of the nation’s nuclear fleet while studying and mitigating aging phenomena and issues that are applicable to the service environments of operating reactors and require unique DOE laboratory expertise. Often, research and development (R&D) and demonstration activities are cost-shared, coordinated, or otherwise jointly conducted with industry, regulatory agencies, or other organizations through the aforementioned or other collaborative mechanisms. Operation of the existing fleet of plants to 60 years, extending the operating lifetimes of those plants beyond 60 years and, where practical, making further improvements in their productivity are essential to support the nation’s energy needs. Recently, several utilities have submitted or announced plans to submit applications to the U.S. NRC to begin the subsequent license-renewal process, extending the operating license period beyond 60 years from the date of their initial licensing. This marks an important planned milestone in the history of commercial nuclear power operations in the U.S.—one that underscores the long-term dependability of these plant designs and the commitment to their long-term performance by the organizations that operate them. The LWRS Program will continue to work with owner-operators to address the key issues needed to support the technical bases for continued safe long-term operation of our nation’s nuclear power assets.

42 ENGINEERING↗
DOE OSTI2018

AGR-2 Irradiation Test Final As-Run Report, Rev 3

This document presents the as run analysis of the AGR 2 irradiation experiment. AGR 2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) program. The objectives of the AGR 2 experiment are to: 1. Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR 1 test and other project activities. 2. Provide irradiated fuel samples for post irradiation experiment (PIE) and safety testing. 3. Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

AGR-2 Irradiation Test Final As-Run Report (Revision 4)

This document presents the as run analysis of the AGR 2 irradiation experiment. AGR 2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) program. The objectives of the AGR 2 experiment are to: 1. Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR 1 test and other project activities. 2. Provide irradiated fuel samples for post irradiation experiment (PIE) and safety testing. 3. Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

AGR-2 Irradiation Test Final As-Run Report, Rev 2

This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO 2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO 2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO 2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test contained six independently controlled and monitored capsules. Each U.S. capsule contained 12 compacts of either UCO or UO2 AGR coated fuel. No fuel particles failed during the AGR-2 irradiation. Final burnup values on a per compact basis ranged from 7.26 to 13.15% FIMA (fissions per initial heavy-metal atom) for UCO fuel, and 9.01 to 10.69% FIMA for UO 2 fuel, while fast fluence values ranged from 1.94 to 3.47 x 10 25 n/m 2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.53 x 10 25 n/m 2 (E >0.18 MeV) for UO 2 fuel. Time-average volume-average (TAVA) temperatures on a capsule basis at the end of irradiation ranged from 987°C in Capsule 6 to 1296°C in Capsule 2 for UCO, and from 996 to 1062°C in UO 2 -fueled Capsule 3. By the end of the irradiation, all of the installed thermocouples (TCs) had failed. Fission product release-to-birth (R/B) ratios were quite low. In the UCO capsules, R/B values during the first three cycles were below 10 -6 with the exception of the hotter Capsule 2, in which the R/Bs reached 2 x 10 -6 . In the UO 2 capsule (Capsule 3), the R/B values during the first three cycles were below 10 -7 . R/B values for all following cycles are not reliable due to gas flow and cross talk issues.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗