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At least 757 records · Page 42

Integrated Process Testing of MSR Salt Spill Accidents

Part of the licensing process for new nuclear reactors requires vendors to assess the potential consequences of identified accident scenarios using accident progression modeling. The accident scenario that will likely be evaluated by all molten salt reactor (MSR) developers is a spill of radionuclide-bearing fuel salt onto the reactor containment floor (i.e., a salt spill accident). The development of accident progression models requires experimental data to inform which processes to incorporate, to enable the calculation of parameters to model these processes, and to validate the model predictions. The data should quantify the sensitivities of key processes (e.g., molten salt spreading, heat transfer, containment structure corrosion, radionuclide vaporization, and aerosol generation) towards the initial conditions of the spill, the ambient environment, and the features of the containment. In addition, results from integrated process tests that quantify coupled processes are required to validate systems-level models. This report documents results from integrated process tests conducted on simulated molten salt spill accidents. The generated experimental data simultaneously quantify the heat transfer behavior of the spilled salt, compositional changes to the bulk salt, and the release of surrogate fission products from the spilled salt as aerosol particles. All tests that were conducted used FLiNaK doped with surrogate fission products, and the variables that were evaluated included the initial salt temperature and the concentration of surrogate fission products present in the salt. The major accomplishments of this work include identifying surrogate fuel salt compositions that provide insight into the dispersal behavior of radionuclides of potential significance to the source term, employing previously developed methods and measurement techniques to simultaneously measure key processes, generating data on the coupled processes of molten salt heat transfer and surrogate fission product release as aerosol particles, demonstrating new test methods for real-time monitoring of the flow rate of the spill and aerosol size quantification in an argon atmosphere, and developing a mass transfer model for cesium and iodine release from molten FLiNaK to provide insight into aerosol formation by vapor condensation. The same methodology applied herein can be employed to study different salt compositions of interest to MSR developers, different environmental conditions, and other variables that are relevant to postulated accident scenarios. The insights gained from these integrated process tests conducted at a laboratory scale will be incorporated into future integral effects tests conducted at an engineering scale.

20 FOSSIL-FUELED POWER PLANTS↗

Development of ASTM Standard for SiC-SiC Joint Testing Final Scientific/Technical Report

As the nuclear industry moves to advanced ceramic based materials for cladding and core structural materials for a variety of advanced reactors, new standards and test methods are required for material development and licensing purposes. For example, General Atomics (GA) is actively developing silicon carbide (SiC) based composite cladding (SiC-SiC) for its Energy Multiplier Module (EM2), a high efficiency gas cooled fast reactor. Through DOE funding via the advanced reactor concept program, GA developed a new test method for the nominal joint strength of an endplug sealed to advanced ceramic tubes, Fig. 1-1, at ambient and elevated temperatures called the endplug pushout (EPPO) test. This test utilizes widely available universal mechanical testers coupled with clam shell heaters, and specimen size is relatively small, making it a viable post irradiation test method. The culmination of this effort was a draft of an ASTM test standard that will be submitted for approval to the ASTM C28 ceramic committee. Once the standard has been vetted by the ceramics test community, an industry wide standard methodology to test joined tubular ceramic components will be available for the entire nuclear materials community.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Investigation of the irradiation effects in additively manufactured 316L steel resulting in decreased irradiation assisted stress corrosion cracking susceptibility

While additive manufacturing (AM) is an area of increasing interest to the nuclear industry, there are significant research and development needs prior to successful implementation of additively manufactured components into reactors. AM steel behaves differently than traditional wrought material, as has been observed in numerous mechanical testing studies. The nuclear industry must also understand the material response to irradiation. This work examines the irradiation damage and the irradiation assisted stress corrosion cracking behavior (IASCC) of both proton irradiated AM and wrought 316L steel. For IASCC testing, specimens were strained to 4% plastic strain in a simulated boiling water reactor environment (288°C, 0.2 µS/cm). Transmission and scanning electron microscopy, confocal microscopy, and X-ray computed tomography were used to examine coupons of the material as well as tensile bars used in the IASCC testing. This work found that the wrought 316L was more susceptible to IASCC than the AM steel, however, some IASCC was observed in the AM steel when tested with the tensile axis parallel to the build direction. Electron microscopy has shown significantly fewer radiation induced voids formed in the AM steel. While the pre-existing pore network in the AM steel may act as a sink to point defects, this is not expected to be the major factor in the reduced swelling observed in the AM steel. The lower density of radiation induced voids was expected to contribute to the IASCC resistance by decreasing the localization of deformation in the irradiated specimens, however, severity of localized deformation in AM steel, measured in dislocation channel spacing and height, was similar or more severe to that found in wrought. In conclusion, the presence of severe cracking in the wrought specimens did not allow for a direct comparison of localized deformation severities between the AM and wrought steel.

Additive manufacturing↗

Molten salt to supercritical CO2 diffusion-bonded heat exchanger testing to support component certification for advanced nuclear power systems

Commercial utilization of the diffusion-bonded heat exchanger (DBHE) for high-temperature applications has prompted an ASME Boiler and Pressure Vessel Code Case development effort to employ these components in Section III (nuclear plant) applications. As part of this effort, a series of DBHE have been fabricated. These heat exchangers will be among the first printed-circuit geometries to be tested for functionality in a molten salt environment. These 15-25kW test specimens will be evaluated under steady-state conditions to compare pressure drop and terminal temperatures measurements with those predicted by existing flow and heat transfer correlations. The test units will then be subjected to thermal transients similar to those expected in Generation IV reactor systems by varying the hot and cold fluid mass flow rates and by changing inlet temperatures. Data collected from these tests will narrow the technical gaps related to two-fluid PCHE and identify special design requirements of compact heat exchangers containing primary reactor coolant. Descriptions of one of two test units and the testing plan are discussed here.

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Thermal hydraulic analysis of the AFIP-7 irradiation test in the Advanced test Reactor – Model correlation and performance evaluation

We report as a system level code, RELAP5 is widely used in the nuclear field for the reactor hydraulic analysis. In the component/experimental level, it is frequently employed as well. This paper demonstrates that the RELAP5 flow simulation of the AFIP-7 irradiation experiment which has a complicated 3D geometry/flow path deviates by ~28% from the flow evaluation through a computational fluid dynamics (CFD) simulation, which has been validated against a flow test performed at Oregon State University. Thermal safety compliance analysis of an experiment planned to be irradiated in the Advanced Test Reactor is usually performed using the finite element analysis code, Abaqus. This research reveals the disadvantages of the Abaqus simulation in the flow instability and departure from nucleate boiling evaluations via a conjugate heat transfer analysis of the AFIP-7 irradiation experiment. As a result, a detailed CFD simulation is suggested for irradiation experiments with complicated flow paths, rather than a simplified RELAP5 simulation or hand calculations. The validated simulation approach will be integrated with the Boehmite correlations to investigate the oxide growth prediction on the fuel cladding in Part II of this research.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

National Reactor Innovation Center Annual Report FY 2023

It is my privilege to present this year’s annual report for the National Reactor Innovation Center (NRIC). Now rounding out our fourth year of operation, NRIC is poised for successfully delivering our most foundational promise: to accelerate the testing and demonstration of advanced nuclear technology by providing access to national laboratory assets and expertise. Some specific goals include: • Establish and maintain four new experimental facility capabilities: o MSTEC – 2025 o METL – Operational o HeCTF – Operational o Creep Frames – 2026 • Construct and operationalize two advanced reactor test beds (DOME and LOTUS) for integrated technology demonstrations and experimentation – 2028 • Complete minimum of two advanced nuclear technology tests – 2030 o Advanced Construction Technology – 2025 o Advanced Microreactor test in DOME or LOTUS – 2027 & 2029

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Improving to the neutron fluence rate monitor measurement system at the Advanced Test Reactor

The existing fluence monitor wire scanning system at the Advanced Test Reactor (ATR) was designed and installed for use in the Engineering Test Reactor (ETR) when it began operation in 1958. The wire scanner was operated in ETR for over 20 years until ATR began operation, when it was moved to the ATR west canal area in 1971 and subsequently moved to the west canal in 2006 where it presently resides. With a continued service life of 65 years the system is well beyond the typical design life of 20 years for these types of systems. The need to update the data acquisition and control system was identified, and the benefits of replacing the existing sodium iodide (NaI) detector with an electronically cooled high-purity germanium (HPGe) detector are discussed.

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Intracore Natural Circulation Study in the High Temperature Test Facility

The development of the Modular High-Temperature Gas-Cooled Reactor is a significant milestone in advanced nuclear reactor technology. One of the concerns for the reactor’s safe operation is the effects of a loss-of-flow accident (LOFA) where the coolant circulators are tripped, and forced coolant flow through the core is lost. Depending on the steam generator placement, loop or intracore natural circulation develops to help transfer heat from the core to the reactor cavity, cooling system. This paper investigates the fundamental physical phenomena associated with intracore coolant natural circulation flow in a one-sixth Computational Fluid Dynamics (CFD) model of the Oregon State University High Temperature Test Facility (OSU HTTF) following a loss-of-flow accident transient. This study employs conjugate heat transfer and steady-state flow along with an SST k-ω turbulence model to characterize the phenomenon of core channel-to-channel natural convection. Previous studies have revealed the importance of complex flow distribution in the inlet and outlet plenums with the potential to generate hot coolant jets. For this reason, complete upper and lower plenum volumes are included in the analyzed computational domain. CFD results also include parametric studies performed for a mesh sensitivity analysis, generated using the STAR-CCM+ software. The resulting channel axial velocities and flow directions support the test facility scaling analysis and similarity group distortions calculation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Integral and Separate Effects Test Facilities To Support Water Cooled Small Modular Reactors: A Review

This study reviews previous experimental facilities and test programs relevant to water-cooled reactor system design and analysis to meet regulatory compliances. This study aims to find the best solution for designing the required experiments, obtaining necessary test data, and verifying the developed computer code/models to support the new reactor design and development while minimizing cost and time while leveraging experiences from previous facilities to minimize. Nuclear reactor licensing requires supportive design, analysis, and experimental results to ensure the safety of the full-scale prototype reactor in regular operation, as well as during postulated accident scenarios. These reactor design analyses are generally performed using system codes and other associated simulation tools that require assessment, verification and validation using an appropriate experimental dataset. Experimental facilities used for reactor system safety analysis and system code assessments are categorized into integral effects test (IET) and separate effects test (SET) facilities. Further, these IET and SET experiments and studies use geometrically scaled systems to reproduce the prototype system behavior at a reasonable cost, albeit with some scaling-related distortions. The design challenge of these model facilities is to identify and minimize scaling distortions while reproducing the most important operational phenomena in steady-state operation and in postulated accident scenarios. Lessons learned from previous experimental facilities, models, and correlations can support the development of new multipurpose, scaled, hybrid, integrated, and modular experimental facilities for advanced light water-cooled small modular reactors (SMRs). Successful operation of these facilities can significantly reduce upfront reactor development and demonstration costs and time to deployment.

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Initial Gamma Spectrometry Examination of the AGR-3/4 Irradiation - 18594

The initial results from gamma spectrometry examination of the different components from the combined third and fourth US Advanced Gas Reactor Fuel Development TRISO-coated particle fuel irradiation tests (AGR-3/4) have been analyzed. This experiment was designed to provide information about in-pile fission product migration. In each of the 12 capsules, a single stack of four compacts with designed-to-fail particles surrounded by two graphitic diffusion rings (inner and outer) and a graphite sink were irradiated in the Idaho National Laboratory s Advanced Test Reactor. Gamma spectrometry has been used to evaluate the gamma-emitting fission product inventory of compacts from the irradiation and evaluate the burnup of these compacts based on the activity of the radioactive cesium isotopes (Cs-134 and Cs-137) in the compacts. Burnup from gamma spectrometry compares well with predicted burnup from simulations. Additionally, inner and outer rings were also examined by gamma spectrometry both to evaluate the fission product inventory and the distribution of gamma-emitting fission products within the rings using gamma emission computed tomography. The cesium inventory of the scanned rings compares acceptably well with the expected inventory from fission product transport modeling. The inventory of the graphite fission product sinks is also being evaluated by gamma spectrometry. (authors)

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Sensitivity Analysis of an Experimental Breeder Reactor II Fuel Assembly

The Experimental Breeder Reactor II (EBR-II) was selected as a candidate for a reactor physics benchmark evaluation due to its unique features and capabilities. In connection with the benchmark evaluation, the goal of this work was to perform a sensitivity analysis. The EBR-II was a sodium cooled, pool type, metal fueled, fast spectrum reactor. EBR-II operated from 1964 through 1994. It had a maximum heat output of 62.5 MW which could produce 20 MW of electricity. The design of EBR-II consisted of 637 hexagonal shaped, removable assemblies. The assemblies were divided into three regions: the core, an inner blanket, and an outer blanket. The core region contained assemblies with 91 fuel rods in each. The fuel was enriched uranium metal and was clad within stainless steel. The core region also included two safety assemblies and eight control assemblies with the difference being that these assemblies only contained 61 fuel rods. Additional assemblies in the core region included stainless steel dummies, half worth drivers, and experimental/instrumentation assemblies. The inner blanket region initially consisted of depleted uranium for fuel breeding, but was replaced, after proving the concept of breeding, with stainless steel reflectors for the purpose of irradiation. The outer blanket region consisted of depleted uranium. Figure 1 shows a cross section view of the EBR-II core layout. Each assembly was approximately 234 cm long with the uranium fuel section being 33 cm long. Above and below the fuel were stainless steel neutron reflectors. Additionally, orifices at the bottom of the assembly allowed for sodium coolant to flow upward, with larger holes in the subassemblies at the center where the greatest heat was produced. The core of EBR-II was enclosed by a stainless steel reactor vessel which was comprised of a radial shell, a lower grid plenum structure, and an upper reactor vessel cover that served as a neutron shield. Surrounding the reactor vessel radially were layers of graphite and borated graphite shielding. The entire assembly is submerged 3 meters in a pool of sodium. The pool contained up to 337,000 liters of sodium which was forced through the core and blanket regions using two centrifugal pumps that were also submerged within the sodium pool. The design of EBR-II included many safety features. The compatibility between the sodium and stainless steel prevented corrosion and therefore the risk of a radioactive release due to corrosion. The sodium coolant was kept at near atmospheric pressure which made any leak more controllable. The metal fuel, cladding, and coolant allowed for heat to be removed easily. The large volume of sodium allowed for heat to be absorbed without problems occurring. Lastly, the reactor would expand as temperature increased which would result in greater neutron leakage. This extra leakage caused a large negative temperature reactivity feedback which automatically shut the reactor down in case of an accident. EBR-II had unique capabilities which were proven during the Integral Fast Reactor initiative. This program lasted from 1984 to 1994. The goal of this program was to solve many of the perceived problems with nuclear energy through scientific means. The safety features of EBR-II were put to the test in April 1986. The tests simulated loss of reactor coolant flow and loss of a heat sink. For both of these tests, the reactor was at 100% power and was not scrammed. During both tests, the reactor shut itself down and achieved a safe steady state temperature all due to the design of the reactor. It was because of these tests that EBR-II was chosen for a reactor physics benchmark evaluation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

KRUSTY Test Results

Testing of the Kilopower Reactor Using Stirling Technology (KRUSTY) was conducted in 4 phases over 5 months: from November 2017 to March 2018, at the Nevada National Security Site (NNSS). This report contains a summary of test results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

SULI Internship Deliverables (Flow Orifice Design: Measuring Degas Flow of Experiment Loops at the Advanced Test Reactor)

These deliverables present a design for a new orifice flow meter to be installed in three experiment loops at the Advanced Test Reactor. The current configuration of the experiment loop system has an orifice that is not properly sized to the low flow rates of the degas line of the pressurizer system, causing inaccurate flow measurements. Under consideration of flow conditions and system constraints, an iterative procedure was utilized to determine the optimal sizing for a new orifice. This orifice is designed to accurately and precisely measure the flow rate of the degas line, which will consequently improve control of the pressure and chemistry of the loop pressurizer system. The scope of this project includes recommendations for orifice sizing, material, flange type, face type, and installation location on the loop piping spools. The design is suitable for experiment loops 1D-N, 2B-SE, and 2D-SW and meets all necessary standards.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Optimization of a dry, mixed nuclear fuel storage array for nuclear criticality safety - 14339

A dry storage array of used nuclear fuel at the Idaho National Laboratory contains a mixture of more than twenty different types of research and test reactor fuels. Significant neutronic interaction occurs between fuel canisters within this large storage arrangement of 636 fuel storage canisters. Nuclear criticality safety is maintained primarily by moderation control and through control of neutron interaction and leakage provided by the credited geometry. Over the nearly forty years of operation the collected fuels are now arranged in a configuration that does not minimize the system neutron multiplication factor (k{sub eff}). This creates a concern since the entire remaining facility storage capacity cannot be utilized without exceeding the system minimum subcritical state (k{sub safe}) for ensuring nuclear criticality safety. This work determines a more optimal arrangement for the stored fuels with a goal to minimize the system k{sub eff}, but also with a minimum of potential fuel canister relocation movements. The ultimate goal is to establish a solution for this multiple-objective optimization problem by finding the Pareto Front for maximum k{sub eff} reduction versus the number of canister relocations. The process first required additional analysis of the array by Monte-Carlo neutron transport codes (specifically, MCNP) to gain a greater understanding of the interaction characteristics of the disparate fuel types. Some stored fuels are composed of primarily metallic and oxide components that provide very little inherent moderation, while other fuel types contain graphite or zirconium-hydride which can contribute a significant amount of moderation. These factors cause a complex relationship within a fuel arrangement with respect to thermal utilization and axial leakage probability. The new MCNP analysis was utilized to develop a simplified spreadsheet-based empirical model which simulates the array k{sub eff} result. This empirical model was benchmarked against additional MCNP calculations for various arrangements and then employed in the optimization analysis. A Tabu metaheuristic and stochastic approximation optimization was applied to the empirical model with the optimization function to minimize the array k{sub eff}. The optimization algorithm interrogated the empirical model to determine the optimal rearrangement for up to sixty canister movements. The Pareto Front was found, which shows that a limited number of canister relocation movements could allow the existing thirty-one empty array positions to be filled with common research reactor fuels while also yielding a reduction in the overall array k{sub eff} relative to the current arrangement. This will impart an improvement in the facility utilization while also offering an enhancement in the safety margin. The reduction in array reactivity has been confirmed by a final set of MCNP calculations with the proposed new arrangements. (authors)

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Regulatory Approach for Nuclear Thermal Propulsion Reactor Systems

Nuclear Thermal Propulsion (NTP) is being developed to support crewed or cargo transfer missions to Mars. Obtaining regulatory approval for NTP engine testing proves a challenge for not only the technology maturation required but also the fabrication, launch and operational nuclear regulatory environments. Existing regulations currently apply to either (1) high power, long duration commercial power plants, or (2) low power, short duration research reactors. NTP systems find themselves in a separate and unique area due to their higher power but short operating duration. This is supplemented by use of the reactor coolant as propellant. When planning the regulatory approach for NTP systems, operations in space and testing on Earth must be considered – each have their own challenges. This paper identifies a preliminary pathway for regulatory approval to operate a NTP demonstration engine as well as supporting test data that is recommended to be generated during the development program to support major regulatory milestones and deliverables.

Regulatory↗

Reactor Parametric Assessments for Alternative Propellant Nuclear Thermal Propulsion Engines

This work focuses on the implications of alternative working fluids for nuclear thermal propulsion (NTP) reactors. To perform this analysis, NASA’s Testing Reference Design (TRD) is altered using parametric studies with selected alternative propellants to determine the reactor modifications required to enable an operable system. This research investigates ammonia, water, methane, helium, and enriched diborane as feasible options to enable a specific impulse (Isp) greater than the chemical propulsion alternatives. Frozen and dissociated Isp for each propellant is shown for variable chamber temperature conditions. Geometrical sensitivities are performed to observe the impact on the neutron multiplication factor (keff), system mass, and HALEU loading. Control drum worth curves are included for 5- degree increments of rotation. The base TRD configuration does not enable a critical system for ammonia and enriched diborane, thus these propellants will be highlighted in the final results. Equivalent study results for hydrogen will be included as a point of comparison to the performance of the other options.

Nuclear Thermal Propulsion↗

Parametric Study of Used Nuclear Oxide Fuel Constituent Dissolution in Molten LiCl-KCl-UCl 3

Prior work identified dissolution of used nuclear oxide fuel constituents from a uranium oxide matrix into molten LiCl-KCl-UCl 3 at 500°C, prompting a subsequent series of three progressive studies (including an initial scoping study, an electrolytic dissolution study, and a chemical-seeded dissolution study) to further investigate associated parameters and mechanisms. Thermodynamic calculations were performed to identify possible reaction mechanisms and their propensities in used oxide fuel constituent dissolution. Used nuclear oxide fuels with varying preconditions from fast and thermal test reactors were separately immersed in the subject salt system to assess fuel constituent migration from the bulk fuel matrix to the salt phase in an initial scoping study. Dissolution of expected fuel constituents, including alkali, alkaline earth, lanthanide, and transuranium oxides, into the chloride salt phase varied widely, ranging from 12% to 99% in the initial study. Uranium isotope blending between the salt phase and bulk fuel matrix was also observed, which was attributed to reducing conditions in the fuel matrix. Electrolytic and chemical-seeded dissolution studies were subsequently performed to effect reducing conditions in the fuel. Other parameters, including temperature (at 500°C, 650°C, 725°C, and 800°C) and uranium trichloride concentrations (at 6, 9, and 19 wt% uranium), were investigated in the latter two studies, resulting in fuel constituent dissolution above 90%. Extents of dissolution were based on initial and final fuel constituent concentrations in the oxide fuels following operations in the salt and subsequent removal of the salt via distillation. Finally, in this series of progressive studies, oxide fuel preconditioning and in situ reducing conditions, along with elevated temperature and uranium trichloride concentrations, were the primary parameters promoting used nuclear oxide fuel constituent dissolution in accordance with identified reaction mechanisms.

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