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At least 847 records · Page 47

Assessment of Microreactor Safety Analysis Challenges and Recommendations for Utilization of the Comprehensive Reactor Analysis Bundle

To enable the broad deployment of microreactors in fundamentally new application regimes (i.e., mobile and autonomous operations), their safety must be indisputable in terms of possessing inherent resistance to severe offsite dose consequences. Therefore, mechanistic beyond-design-basis event source term calculations that demonstrate a sufficiently large margin of safety will be required to accommodate these new application regimes, which have no history of commercial regulation. Even for traditional reactor operation configurations, safety analysis expertise and familiarity with accident phenomena and conditions in microreactors—specifically those with heat pipe primary cooling arrangements—are lacking compared with other advanced reactor concepts and small modular reactors. Recently, modeling and simulation tools to account for unique heat pipe design aspects have been developed by Sandia National Laboratories with MELCOR and by the US Department of Energy’s (DOE’s) Office of Nuclear Energy Advanced Modeling and Simulation Program with BlueCRAB. However, further demonstration and assessment of potential knowledge gaps are needed to support these codes’ broad usage by the microreactor community. Through the DOE Microreactor Program, an initial assessment of these two tools and guidance on how an evaluation model could be constructed was performed and is reported herein. Moreover, a proposed approach for demonstrating an evaluation model using these two tools is outlined.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Toward the implementation of self-powered, wireless, real-time reactor power sensing

The successful development of real-time, self-powered, wireless reactor power sensing technology would enable key operating parameters of every fuel assembly in a commercial reactor core to be continuously monitored, and potentially be utilized for applications such as closed-loop reactor control. This paper discusses the expansion of successful work on the application of Self Powered Neutron Detectors (SPND) for reactor power monitoring and the wireless transmission of these detector signals using advanced vacuum micro-electronic (VME) transmitter technology developed by Westinghouse in collaboration with the Penn State University Radiation Science and Engineering Center. Results demonstrate that this technology may be successfully applied via long-term testing in the Penn State TRIGA reactor. SPND are inherently slower than, for example, ion chambers due to their dependence on radioactive decay to generate a signal. Additional research developed delay compensation algorithms to accelerate the response of these detectors for use in near-real-time closed-loop reactor control. Initial results indicate that practical signal acceleration can be obtained digitally or with an equivalent analog circuit. Future research will apply an inverse model of the sensor’s balance differential equations to more-effectively accelerate the SPND signal. Use of inverse models will also prove to be adaptive in the sense that time-dependent burn-up of these detectors, due to neutron activation of the detector material, can be modeled and applied in the acceleration algorithm.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Measure this, not that: Optimizing the cost and model-based information content of measurements

Model-based design of experiments (MBDoE) is a powerful framework for selecting and calibrating science-based mathematical models from data. Here, this work extends popular MBDoE workflows by proposing a convex mixed integer (non)linear programming (MINLP) to optimize the selection of measurements. The solver MindtPy is modified to support calculating the D-optimality objective and its gradient via an external package, scipy, using the grey-box module in Pyomo. The new approach is demonstrated in two case studies: estimating highly correlated kinetics from a batch reactor and estimating transport parameters in a large-scale rotary packed bed for CO 2 capture. Both case studies show how examining the Pareto optimal trade-offs between information content measured by A- and D-optimality versus measurement budget offers practical guidance for selecting measurements for scientific experiments.

97 MATHEMATICS AND COMPUTING↗

Accelerating nuclear-integrated data center pursuits in the USA: SWOT analysis, power-thermal management strategies and demonstration plan

Here, this study explores the increasing interest in leveraging nuclear power to meet the escalating energy demands of data centers in the United States (U.S.) by focusing on key factors that contribute to accelerated deployment. The study highlights the importance of N+1/N+2 power supplies (where N is the required number of units), outlines research and innovations in nuclear-integrated data center thermal management and demonstration plan. It also provides updates about status and costing of various reactor system designs. A summarized strengths, weaknesses, opportunities, and threats (SWOT) analysis shows the potential options for grid connectivity, reactors, and site selection. Suitable site discussions consider land and water availability, grid access, and optical fiber connectivity, and the study presents graded prospects for Department of Energy (DOE) sites with a specific example. Community engagement and partnerships are emphasized, particularly the roles of local government, federal agencies, utilities, and data center industry partners, which are crucial for accelerating deployment, business outreach, and approvals. The study provides actionable insights for stakeholders to accelerate the deployment of nuclear-powered data centers.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

Controls on Barite (BaSO 4 ) Precipitation in Unconventional Reservoirs

Barite (BaSO 4 ) precipitation is one of the most ubiquitous examples of secondary sulfate mineral scaling in shale oil and gas reservoirs. Often, a suite of chemical additives is used during fracturing operations to inhibit the accumulation of mineral scales, though their efficacy is widely varied and poorly understood. This study combines experimental data and multi-component numerical reactive transport modeling to offer a more comprehensive understanding of the geochemical behavior of barite accumulation in shale matrices under conditions typical of fracturing operations. A variety of additives and conditions are individually tested in batch reactor experiments to identify the factors controlling barite precipitation. Our experimental results demonstrate a pH dependence in the rate of barite precipitation, which we use to develop a predictive model including a pH-dependent term that satisfactorily reproduces our observations. Further, this model is then extended to consider the behavior of three major shale samples of highly variable mineralogy (Eagle Ford, Marcellus, and Barnett). This data-validated model offers a reliable tool to predict and ultimately mitigate against secondary mineral accumulation in unconventional shale reservoirs.

54 ENVIRONMENTAL SCIENCES↗

High temperature gamma radiation-induced chromium redox chemistry via in situ spectroscopic measurements

Chromium ions can make their way into the primary coolant of nuclear power reactors from the corrosion of stainless-steel reactor components, decreasing the material's corrosion resistance and resulting in increased transport of further corrosion products. Despite these potential effects, the radiation-induced redox speciation of chromium ions in aqueous solution is not well understood, especially at the elevated temperatures experienced by reactor coolants. Here, in the present work, we report new experimental results demonstrating that in aerated aqueous solution, the radiolytic oxidation of Cr(III) to Cr(VI) occurs at pH 4, while the reduction of Cr(VI) to Cr(III) occurs at pH 2. The oxidation of Cr(III) is primarily attributed to the reaction of the hydroxyl radical (˙OH) with the Cr(OH) 2+ species, while the reduction of Cr(VI) is attributed to reactions involving the hydrated electron (e aq - ) and hydrogen atom (H˙). Additionally, the steady-state equilibrium yield of Cr(VI) from the gamma irradiation of pH 4 Cr(III) solutions decreased with increasing temperature (over a range of 37–195 °C). This observation indicates that the activation energy of the Cr(VI) reduction reactions is higher than that for the Cr(III) oxidation reactions, such that it becomes relatively more favorable at higher temperatures. Overall, these data are important for the development of complementary multiscale models for the prediction of metal ion speciation in high temperature radiation environments.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Demonstration of tokamak vertical stability control based on non-inductive Faraday-effect polarimetry measurements

Long-pulse or steady-state fusion reactors are envisioned to control vertical stability based on non-inductive measurements, i.e. that do not rely on temporal change of magnetic field. For the first time, vertical stability control using non-inductive Faraday-effect polarimetry measurements has been demonstrated. The Radial Interferometer-Polarimeter system on DIII-D is capable of microsecond resolution and was used to absolutely determine the vertical position of the plasma magnetic axis Z 0 . A vertical stability controller was developed to robustly stabilize diverted plasmas using Faraday-based measurements. The system was able to stabilize against vertical displacement events with growth rates up to 350 s -1 in elongated and elliptical plasma shapes, and instabilities with even higher growth rates are likely to be controllable with further improvements to controller tuning. Tests show that the Faraday-based controller remains effective and is capable of recovering from loss of control even when the plasma vertical position is far from the region where the linear model used to calculate Z 0 is most valid. Faraday control has also been activated during plasma ramp-up, demonstrating the robustness of the technique to larger systematic diagnostic uncertainty at low electron density.

Faraday-effect polarimetry↗

Materials and Fuels Complex Five-Year Investment Strategy (FY2020-FY2024)

The Department of Energy Office of Nuclear Energy (DOE-NE) mission is to provide the research, development, and demonstration (RD&D) foundation to extend the lives of the current operating reactor fleet, develop the next generation of nuclear reactors, and provide integrated nuclear fuel cycle solutions. The Materials and Fuels Complex (MFC) serves as the foundation of a nuclear RD&D enabling test bed at Idaho National Laboratory (INL) and is an integral part of a National Reactor Innovation Center (NRIC) strategy. MFC facilities focus on developing and maintaining RD&D capabilities that can increase research throughput, reduce barriers to deployment, and facilitate commercialization of new ideas and technologies for clean and secure sources of energy. This document is a complimentary document to the Materials and Fuels Complex Five-Year Mission Strategy FY-20 – FY-24 (INL/EXT-20-57224) that defines specific implementation strategies for increasing research capability and throughput through targeted investments in research facilities, research instruments, and research staff. Refer to INL/EXT-20-57224 for details about MFC, its capabilities, and the overall mission strategy for the Materials and Fuels Complex.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Radio Frequency Field Programable Gate Array Implementation of Reflectometry Cable Monitoring

This document describes the development of a field programable gate array (FPGA) radio frequency system on a chip (RF SoC) adaptation and evaluation of the single-board device to perform both Frequency Domain Reflectometry (FDR) and Spread Spectrum Time Domain Reflectometry (SSTDR) for offline and online cable testing. The work builds on and leverages the work of Pacific Northwest National Laboratory (PNNL) in airport millimeter wave technology by using the same development hardware employed in that program. The work is performed under sponsorship from the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program and the task objective is to confirm and demonstrate feasibility to adapt FPGA technology for a cost-effective multiplexed single-board electronic module to perform cable tests that are equivalent to commercial and laboratory test instruments for FDR and SSTDR cable tests. The developed 2-channel (extendable to 7 channels) system was compared to dedicated and proven test instruments and shown to produce equivalent results on a range of cables and with a range of damage types. The FPGA reflectometry test board is one of several technologies that could facilitate implementation of online monitoring of safety critical cable systems.

42 ENGINEERING↗

Progress Towards Synthesis of Uranium Chloride Fuel Salts Using Zinc Chloride

Reliable, scalable methods for producing high-purity actinide chloride salts are needed to support molten salt reactor fuel development and deployment. This report describes the continued development and demonstration of a bench-scale chlorination and purification apparatus using a zinc chloride-based method for synthesizing uranium chloride fuel salts. In this approach, uranium metal is chlorinated by ZnCl2 to produce LiCl-KCl-UCl3. Reaction with three aliquots of added uranium metal was used to generate a target uranium concentration of 30 wt %. While this concentration was chosen for initial testing of the apparatus and method, the final uranium concentration is not limited to 30 wt %. The zinc metal generated in the reaction forms an immiscible layer that was removed by volatilization at moderately high temperatures. Electrochemical measurements confirmed the removal of zinc and applied sensing methods indicated the uranium concentration to be approximately 25 wt %. These initial results demonstrate that the bench-scale chlorination apparatus is an effective platform for the synthesis and purification of uranium chloride salts using ZnCl2. This method shows promise for application to industry-relevant salt systems such as NaCl-UCl3. Further development is recommended to optimize reagent loading, zinc removal, and avoid possible U-Zn alloy formation.

Dulovic, Stephanie↗

Testing piezoelectric sensors in a nuclear reactor environment

Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE’s FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this work piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65x1020 nf/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 x1020 nf/cm2, Zinc Oxide is capable of transduction up to at least 6.27 x1020 nf/cm2 , and Aluminum Nitride is capable of transduction up to at least 8.65x x1020 nf/cm2.

T. Reinhardt, Brian↗

National Space Transportation Systems Program mission report

The 515-41B National Space Transportation Systems Program Mission Report contains a summary of the major activities and accomplishments of the sixth operational Shuttle flight and fourth flight of the OV-099 vehicle, Challenger. Since this flight was the first to land at Kennedy Space Center, the vehicle was towed directly to the OPF (Orbiter Processing Facility) where preparations for flight STS-41C, scheduled for early April 1984, began immediately. The significant problems that occurred during STS-41B are summarized and a problem tracking list that is a complete list of all problems that occurred during the flight is given. None of the problems will affect the STS 41C flight. The major objectives of flight STS-41B were to successfully deploy the Westar satellite and the Indonesian Communications Satellite-B2 (PALAPA-B2); to evaluate the MMU (Manned Maneuvering Unit) support for EVA (Extravehicular Activities); to exercise the MFR (Manipulator Foot Restraint); to demonstrate a closed loop rendezvous; and to operate the M.R (Monodisperse Latex Reactor), the ACES (Acoustic Containerless Experiment System) and the IEF (Isoelectric Focusing) in cabin experiments; and to obtain photographs with the Cinema 360 Cameras.

Collins, M. A., Jr.↗

Computing Properties Of Pure And Mixed Fluids

GASPLUS created as two-part code: first designed for use with pure fluids and second designed for use with mixtures of fluids and phases. Offers routines for mathematical modeling of conditions of fluids in pumps, turbines, compressors and other machines. Other routines for calculating performance of para/ortho-hydrogen reactor and heat of para/normal-hydrogen reaction as well as unique convergence routine demonstrates engineering flavor of GASPLUS. Written in FORTRAN 77.

Fowler, J. R.↗

High Fidelity Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Results

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronie response provides a bridge between electrically heated testing and fueled nuclear testing, providing a better assessment of system integration issues, characterization of integrated system response times and response characteristics, and assessment of potential design improvements' at a relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design can developed. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

Bragg-Sitton, Shannon M.↗

High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Test Results

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer. and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics and assess potential design improvements at relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design is developed: this is followed by engineering design, fabrication, and testing to validate the overall design process. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

Bragg-Sitton, Shannon M.↗

Acoustic Monitoring of Pyroprocessing Equipment

This paper provides an introduction to using acoustic monitoring to advance detection techniques for pyroprocessing in support of nuclear safeguards and non-proliferation. The usage of free air acoustic monitoring has been previously demonstrated at Idaho National Laboratory (INL) facilities such as the Advanced Test Reactor and the National Security Test Range. However, the proposed work revolves around a new deployment environment, the Fuel Conditioning Facility, that brings forward several questions regarding the performance of the technology in non-free air media. The confinement of the pyroprocessing equipment to a heavily shielded hot cell, the atmosphere of the hot cell containing argon gas, and the radiation dose inside the hot cell are all new environments for acoustic monitoring. To our knowledge, acoustic measurements have not been completed in such an environment before. This offers a new opportunity to study not only the acoustic signatures of the equipment inside of the hot cell, but also the propagation of the signals through the hot cell and at distances away from their origination. The objective of this paper is to explain the planned instruments to monitor the Fuel Conditioning Facility in order to evaluate acoustic signals emitted from equipment during various stages of operation. Identifying these signals can potentially enable the identification of specific pieces of equipment used in pyroprocessing and produce information of their operational status. If successful, this type of monitoring could offer a new method to aid in safeguards and proliferation detection of pyroprocessing activities.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Using Resin to Reduce Silica in Borated Fluid - 20072

The Electric Power Research Institute (EPRI) strongly recommends the concentration of silica in the reactor coolant be limited, potential ingress be carefully controlled, and frequently monitored to assess potential silica deposits on fuel cladding surfaces. Boraflex is a product with silicone rubber encasing a neutron absorber for criticality control, and it was installed in the majority of nuclear power plant spent fuel pools. Over time and in the presence of radiation, the silicone rubber lining degrades into silica (SiO{sub 2}) and hydrogen gas. Silica is somewhat soluble in water and spent fuel pool water communicates with other plant systems during refueling outages, leading to a buildup of silica in the reactor coolant and connected systems. Due to the prevalence of Boraflex racks in spent fuel pools, the presence of silica is a chronic nuclear industry concern. To address this concern, many nuclear plants have installed reverse osmosis skid systems to remove the silica from their spent fuel pools and, by extension, from reactor coolant. These skids are expensive to install and maintain. Arcadis personnel demonstrated that an iron-impregnated resin could be used to remove silica from borated systems without impacting the water quality. The resin was readily available, inexpensive, and easy to install in existing plant equipment. During a nuclear power plant's refueling outage the iron-impregnated resin proved to be effective in reducing the need for feed and bleed operations, with concurrent cost savings and reduction in required outage water management activities. Thus, the following savings were realized: - Eliminated the need to purchase, install, and maintain a reverse osmosis system avoiding more than $10 million in up-front capital cost and ongoing operating and maintenance costs. - Removed 9.57 kilograms (21 pounds) of silica dioxide from the primary system fluid. - Reduced the volume of feed and bleed from 70,000 gallons (2.65 E+05 liters) to 35,000 gallons (1.32 E+05 liters). - Saved $380,000 in water production and disposal costs. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

USABILITY EXTENSION OF THE ŠKODA VPVR/M CASK FOR TRANSPORTING IRRADIATED FUEL ASSEMBLIES

New abstract from the final version being submitted now: This paper discusses the evolution and enhanced usability of the ŠKODA VPVR/M cask for the transport of irradiated fuel assemblies, particularly within the context of the demand for the delivery of newly appearing irradiated HEU fuel types for which the cask did not yet have a license to transport. Over time, the cask’s internal basket construction has demonstrated notable adaptability to accommodate various exotic HEU fuel types from research reactors of differing origins. The paper outlines sever-al custom internal baskets developed for specific fuel types, including those from Belarus and Serbia, as well as from Georgia, Uzbekistan, and MNSR cores, as well as a recently designed basket for MTR and TRIGA assemblies. The findings high-light the high flexibility and adaptability of the cask, supported by successful rede-signs and licensing efforts, underscoring its value for the safe and secure transport of nuclear material. Old abstract from the draft version that was already approved: The Russian Research Reactor Fuel Return (RRRFR) program, since its inception, has continuously used the ŠKODA VPVR/M Cask fleet designed for the repatriation of irradiated highly enriched uranium (HEU) fuel. As the program progressed (from shut-down and a quasi-abandoned reactor, and/or as it began to include fuels of Chinese and US origin), new challenges emerged for the transport Cask. These were fuel types that had not yet been licensed for the Cask. Although these requirements did not arise during the design of the basic ŠKODA VPVR/M Cask, as revealed by the retrospective analyses, the Internal Basket of the ŠKODA VPVR/M Cask gives a high degree of flexibility to accommodate additional fuel types. This paper provides a brief overview of the ŠKODA VPVR/M Cask, which holds a B(U) type license, and introduces the different types of Internal Baskets that have already been licensed to transport so-called exotic irradiated HEU fuel types, in addition to the original license. The paper presents a new Internal Basket design for accommodating MTR-type and TRIGA-type irradiated HEU fuel assemblies. This includes a detailed presentation of the design basis and the new MTR-TRIGA Internal Basket, as well as the licensing matters of the package under the name ŠKODA MTR-TRIGA Cask, and the conformity test (dry- and wet-run) operations made to verify compliance with the new Internal Basket. Then, as a summary, the usage record for the Cask fleet is presented, and finally, the paper concludes with the consolidated experiences gained during the utilization of the ŠKODA VPVR/M Cask fleet, emphasizing the high degree of Cask flexibility ensured by the Internal Basket’s construction.

42 - ENGINEERING↗